ENVIRONMENTAL ANALYSIS
OF THE URANIUM  FUEL CYCLE
 PART  III ^Nuclear Fuel Reprocessing
                October 1973
 U..S. ENVIRONMENTAL PROTECTION AGENCY
          Criteria and Standards Division
            Washington, D.C. 20460

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            KllVIKONHKIM'AI. ANALYSIS  Of' Ti'K IIHAIIIUit FHKI, CYUi;

                  PART m -  HU(;i,i',Ais  .I'ur.i. KKPUiM'.KSsi)•«;



                                 KKKATA
Paf.e 9., line  1H  - "annual   role.s"  should road "max iimim annual, doses"
Pant: 9, line  20  - detet.e "both  annual  and"
P.if.R 9, line  2.0 t- "tloso.s" Hlnnild  read  "dose roimn.! t.ruenl:r>"

P«ip,(» .10, table 2, caption l:.it:.1c -  "Poinilation DOKC" should re;id "'Population
   Dose Commitment Inc.reinnnl.'s"
Pago 10, Lnhje 2, th.i rd column -  add  t:hc caption "Maximum Individual DOKC"
Page 30, table 2, fourth thru sixth  columns - (he c.ncion  should rend
   "Average Orp,an Do.se Cloiniti.itme.ivt  from an Annual Release  (J9BO startup)"';
   the units  are  "ninn-rein", not "mnn-vem/yr"
Page 10, table 2, fiftli 'column  -  "U.S." should read "rest of U.S."
                                                      9
Pago J.A, figure  1 - the. I)F for  Pu-239  should read 10

Pages 15-18,  figures 2, 3, and  /i  - the curves labeled  "future" represent
   the sum.of health effects induced  by both future and past exposures.

Page C-16, line  23 -.add after  "individuals", "in a population of 10 "

Page CrH, line  2 - "r.ad" should  read  "30 rads" and add after "year", "at
   a dose  rate of one rad per year"
                     Mini' .,1	1.1  „	)  'iin"n
Page C-21,  line 8 - "10" sliould  read  "10

Page D-17,  last line - t» should read t1

Page D-22  - PW = ACXpwf

Page D-23  - Correct last column  to reflect above

Page 22 -  Correct 3rd and 5th  columns to reflect above,

Figurcr. 6,  7,  l>-2, l)-3 - Correct to reflect above.

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                              FOREWORD
     The generation of electricity by light-water-cooled nuclear power
reactors using enriched uranium for fuel is experiencing rapid growth in
the United States.  This increase in nuclear power reactors will require
similar growth in the other activities that must exist to support these
reactors.  These activities, the sum total of which comprises the uranium
fuel cycle, can be conveniently separated into three parts:  1) the
operations of milling, conversion, enrichment, fuel fabrication and
transportation that convert mined uranium ore into reactor fuel, 2) the
light-water-cooled reactor that burns this fuel, and 3) the reprocessing
of spent fuel after it leaves the reactor.

     This report is one part of a three-part analysis of the impact of
the various operations within the uranium fuel cycle.  The complete
analysis comprises three reports:  The Fuel Supply (Part I), Light-Water
Reactors (Part II), and Fuel Reprocessing (Part III).  High-level waste
disposal operations have not been included in this analysis since these
have no planned discharges to the environment.  Similarly, accidents,
although of potential environmental risk significance, have also not been
included.  Other fuel cycles such as plutonium recycle, plutonium, and
thorium have been excluded.  Insofar as uranium may be used in high-
temperature gas-cooled reactors, this use has also been excluded.

     The principal purposes of the analysis are to project what effects
the total uranium fuel cycle may have on public health and to indicate
where, when, and how standards limiting environmental releases could be
effectively applied to mitigate these effects.  The growth of nuclear
energy has been managed so that environmental contamination is minimal
at the present time; however, the projected growth of this industry and
its anticipated releases of radioactivity to the environment warrant a
careful examination of potential health effects.  Considerable emphasis
has been placed on the long-term health consequences of radioactivity
releases from the various operations, especially in terms of expected
persistence in the environment and for any regional, national or world-
wide migration that may occur.  It is believed that these perspectives
are important in judging the potential impact of radiation-related
activities and should be used in public policy decisions for their
control.

     Comments on this analysis would be appreciated.  These should be
sent to the Director, Criteria and Standards Division of the Office
of Radiation Programs,
                                              W. D. Rowe,
                                       Deputy Assistant Administrator
                                           for Radiation Programs
                               lii

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                            CONTENTS




                                                               Page




FOREWORD.	   iii




INTRODUCTION	    1




EFFLUENT CONTROL TECHNOLOGY	    6




RADIOLOGICAL IMPACT OF A REPRESENTATIVE PLANT	    8




CUMULATIVE RADIOLOGICAL IMPACT OF THE INDUSTRY	   13




ECONOMICS OF HEALTH EFFECTS REDUCTION AND EFFLUENT CONTROL...   20




SUMMARY AND CONCLUSIONS.	   25




REFERENCES	   29









APPENDIXES




A.  Spent Nuclear Fuel Radioactivity Forecasts




B.  Fuel Reprocessing




C.  Radiological Impact of Fuel Reprocessing




D.  Economic Impact Analysis

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                         FIGURES

1.  Estimated Cumulative Environmental Buildup of Radioisotopes from the
    Total Fuel Reprocessing Industry in the United States. ....... .......   14

2.  Estimated Past and Future Health Effects Committed by Tritium
    Releases from the United States Fuel Reprocessing Industry ........ , .   15

3.  Estimated Past and Future Health Effects Committed by Krypton-85
    Releases from the United States Fuel Reprocessing Industry ...... ....   16

4.  Estimated Past and Future Health Effects Committed by Iodine-129
    Releases from the United States Fuel Reprocessing Industry ..........   17

5.  Estimated Past and Future Health Effects Committed by Actinide
    Releases from the United States Fuel Reprocessing Industry ..... .....   18
6.  Risk Reduction vs Control System Cost . ........ ............... ..... , .   23

7.  Control System Cost Effectiveness vs Control System Cost ....'. .. .....   24

    A.I  Estimated Cumulative Environmental Buildup of Tritium from
         the Fuel Reprocessing Industry in the United States . . « .........   A-ll

    A. 2  Estimated Cumulative Environmental Buildup of Krypton-85
         from the Fuel Reprocessing Industry in the United States. . .....   A-12

    A. 3  Estimated Cumulative Environmental Buildup of Iodine-129
         from the Fuel Reprocessing Industry in the United States .......   A-13

    A. 4  Estimated Cumulative Environmental Buildup of Plutonium-239
         from the Fuel Reprocessing Industry in the United States .......   A- 14

    B . 1  Typical Process Flow Schematic ......... ........ ........... .....   B-4

    B.2  Schematic Diagram Showing the Steps Required for the "Zero
         Release" Reprocessing Concept ..................................   B-ll

    D.I  Projected Total United States Population ..... ............ . .....   D-7

    D.2  Risk Reduction (Mortality and Morbidity) vs Cumulative Control
         System Cost . . ............ .... ............ ........ ...... ........   D-25

    D.3  Risk Reduction (Mortality) vs Cumulative Control System Cost  ...   D-26
                                   vi

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                                  . TABLES       .                          PAGE


1.  Site Characteristics of Nuclear Fuel'Reprocessing Plants  	     5

2.  Projections of Average Annual Population Dose from-a 5 tonne/day
    Nuclear Fuel Reprocessing Plant	,.,,	    10

3.  Projections of Total Health Impact from a 5 tonne/day Nuclear
    Fuel Reprocessing Plant.		    12

4.  Summary of Health Effects and Costs of Emission Controls for a
    5 tonne/yr Fuel Reprocessing Plant	    22

    A.I  Estimated U.S. Fuel Reprocessing Requirements	    A-2

    A.2  Representative Quantities of Potentially Significant Fission
         Products in Spent Reactor Fuels	    A-6

    A.3  Representative Quantities of Potentially Significant
         Activation Products in Spent Reactor Fuels	    A-7

    A.4  Representative Quantities of Actinides Present in Spent
         Reactor Fuels	    A-9

    A.5  Estimated Annual Inventories of Selected Nuclides in Spent
         Reactor Fuels-........'.	•	    A-10

    B.I  General Information for Commercial U.S. Nuclear Fuel
         Reprocessing Plants	    B-2

    B.2  Control System Data for Nuclear Fuel Reprocessing (LWR +
         Recycle Fuels) . •	    B-15
                                                    Q e
    B.3  Comparison of Processes for the' Removal of   Kr from
         Dissolver Off-Gas from a Fuel Reprocessing Plant	    B-17

    C.I  Milk Concentrations of Iodine-131 and Iodine-129 from Given
         Input 'Concentration and Corresponding Doses	...,,....,.,...    C-9

    C.2  Actinide Air-Dose Conversion Factors Relative to. Plutonium-239«»    C-ll

    C. 3  Summary of Media-Dose Conversion Factors	• • •«    C-12

    D.I  Factors Used in -the Assessment of Environmental Radiological
         Impact of Fuel Reprocessing	...,.,	    D-19

    D.2  Estimate of the Economics of a 5 MTU/day Reprocessing Plant	    D-21

    D. 3  Control System Costs	•	    D-23
                                    vii

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                   PART III -  NUCLEAR FUEL REPROCESSING





IMTRODUCTIOM




    Economic analyses performed both by the AEC and by  commercial




investors have concluded that the economical generation of electric




power by nuclear plants requires that valuable isotopes of uranium and




Plutonium be recovered from spent reactor fuels for re-use in new fuel




elements.  Recovery of uranium and plutonium involves mechanical




chopping of spent fuel elements into small pieces and placing them in an




acid dissolver to separate the spent fuel from its metal cladding prior




to chemical separation of useful Isotopes from waste products by some




adaptation of the Purex solvent extraction process.  This operation




results in the controlled release of fission products and other




radioactive waste materials which accumulate in the elements during




burnup in the power reactor.  Since  the quantities of these waste




materials are large, considerable care is taken to assure that the fuel




elements maintain integrity through  the cycle in the reactor.  In




essence, therefore, the fuel reprocessing step breaks this carefully




constructed barrier and, as a consequence, represents the main source of




radioactivity from the nuclear power industry which could potentially




enter the environment.




    Many complex technological, environmental, and biological factors




are involved in judging the impact of radioactivity on the environment.




It is the purpose of this analysis to examine these factors with respect




to fuel reprocessing requirements over the next several decades in terms




of the potential public health and environmental risks involved.

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     The main objectives  of  this review are twofold;  (1)  to determine the




 population health risks  and the-cost effectiveness of waste controls




 relating to these risks  for the present and future auclear fuel




 reprocessing industry, and  (2)  to document the data base and techniques




 for assessing the environmental,impact of the fuel reprocessing industry.




 A number of•considerations  are  involved in accomplishing these objectives.




 These include forecasts  of  fuel' reprocessing requirements through the




 year 2020, a detailed analysis  of effluent control systems, the environ-




 mental transport of radionuclides, and considerations of the resultant




 doses and health effects.   The  data base and techniques  used for these




-analyses are presented in detail in the appendixes of this report.




     These studies are based on the performance of a hypothetical




 reprocessing plant having characteristics typical of plants now under




 construction.  This model plant was used to evaluate potential




 reductions in health effects for various controls that limit the release




 -of radionuclides to the  environment.  Control system costs were




 considered as well as .their relation to. total plant costs.  These




 analyse's were then expanded to project the impact of all the nuclear




 fuel reprocessing operations expected in the United States up to the




 year 2020. , Doses and world wide health effects have been estimated on




 the basis of projected nuclear power production and anticipated




 radioactive effluent control .techniques.  Only the contribution from




 United States reprocessing facilities and only the most significant




 radioisotopes, in terms of total effect and persistence in the

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environment, have been considered in this analysis.  Major changes in




normal operating procedures, such as shorter cooling time (storage




between fuel discharge from the reactor and initiation of reprocessing)




should significantly affect local and regional doses but would not




appreciably affect national and world impact.    >



  • . •Three reprocessing plants, with an estimated combined annual




capacity of 2,700 metricjtons of_ju^anium_QMTU), are expected to be in




operation in the United States in the next several years.  The fuel




reprocessing industry is expected to increase along with nuclear power




growth to about 50 to 60 plants with reprocessing capacity of about




80,000 Mill per year projected by the year 2020,




     Up to the present, only one commercial facility has been




operational, but is now shut down for modification.  This is the tfuclear




Fuel Services plant in West Valley, New York, with a capacity of 1 tonne




of uranium per day.  Processing of spent nuclear fuel was initiated in




1966 and continued with sporadic interruptions until December 1971 when




the plant was shut down to permit expansion of processing capability




which will increase plant capacity to 3 tonnes per day.  The plant is




located on a 3,300 acre tract owned by the State of New York in Ashford




Township, Cattaraugus County, New York (Ref. 1).  Buffalo, New York, is




26 miles from the plant and several of its southern suburbs are within a




25 mile radius.  Considerable dairy farming and other agricultural




activities are conducted close to the site.  New York State conducts a




comprehensive monitoring program around the facility including daily raw




milk surveillance (Ref, 2 and 3).  Extensive studies have been conducted

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at th±s facility by EPA and its predecessor organizations (Ref. 4, 5, 6,




and 7).




    Two other plants are under construction.  The Midwest Fuel Recovery




Plant is a tonne per day plant located at Morris, Illinois, adjacent




to the Dresden Nuclear Power Station on privately owned property (Ref.




8).  The General Electric Company, which has received a permit to store




spent fuel and process unirradiated fuel, owns and operates the plant,




Opefations with irradiated fuel from nuclear power reactors will




probably be started during the summer or fall of 1973.




    The closest population center is Joliet, Illinois, 14 miles from the




plant.  Part of the city of Chicago is included within a 50-mile radius




of the plant, thus significantly increasing the local population of




concern.  Offsite concentrations must be evaluated in terms:of a




multiple source since the three Dresden reactors are in close proximity.




    The Barnwell Nuclear Fuel Plant is designed to reprocess 5 tonnes of




uranium fuel per day (1,500 tonne per year), and is located adjacent to




the Savannah River Laboratories (SRL) in Barnwell County, South




Carolina, on privately owned property (Ref. 9).  Construction of this




facility was begun during the spring of 1971, and the projected




completion date is 1975.




    Augusta, Georgia, which- is 31 miles from Barnwell, is the nearest




population "center.  The. population within 50 miles of the facility is



about 500,000 or a factor of 10 below the corresponding population at




Midwest.  Additional site characteristics of these three facilities are




presented in table 1.

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                                   TABLE 1




             Site Characteristics of Nuclear Fuel Reprocessing Plants
'
Capacity
MTU/ day
MTU/year
Population within
10 miles
25 miles
50 miles
Population Center
(10 CFR 100)
Maximum Off-site
Annual Average x/Q
(sec/m^)
(1) Technical Specifications
(2) Adapted from ORNL-4451
NFS
1 (3)
300 (900)
21,409
293,711
Buffalo
(532,759)
26 miles
IxlO"6 (1)
(2.2xlO-7) (2)
for NFS (Ref. 1)
(Ref. 11)
MIDWEST
1
300
25,000
303,460
6,300,000
Joiiefc
(67,000)
14 miles
2.8xlO~8 (3)
(L.lxlO~7) (2)

BARNWELL
5
1,500.
7,100
65,209
60}, 084
Augusta
(70,600) -
31 miles
5.7xlO~8
•-
(3)   Dresden stack use assumed

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                                    6






    The Western Interstate Nuclear Board has been assessing the need for




a fuel reprocessing facility in the Western region.  It has concluded




that such a facility will be required in the late 1970's but may be




profitable during the middle 1970's (Eef. 10).




EFFLUENT CONTROL TECHNOLOGY




    Current practices and effluent control measures in the nuclear fuel




reprocessing industry are not totally indicative of the workload and




performance to be expected in new plants coining online during the period




discussed in this report (1970-2020).  Because newer designs will differ




substantially from the first commercial facility, Nuclear Fuel Services?




presently available performance data are of limited use in predicting




environmental releases.  The two newer plants, the Midwest Fuel Recovery




Plant at Morris, Illinois, and the Barnwell Nuclear Fuel Plant at




Barnwell, South Carolina, are far enough along in their design and




construction to provide some idea of what the cost and expected




performance characteristics of new plants may be.  Both of these plants




will utilize the recycling of liquid wastes so that in theory the only




radioactive liquid discharges will be from spills, accidents, or leaks.




The tritium present in the fuel will be released as a stack gas.  This




approach is in contrast to the NFS facility which uses a system designed




to release low-level radioactive liquid wastes into the environment.




    Waste control systems under construction for use in fuel




reprocessing plants and their costs are described in appendix B.  The




design of NFS and the two plants under construction calls for the




release of radioactive krypton-85 as stack gas, although the owners of

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Midwest and Barnwell have indicated a willingness to add krypton removal


equipment when it is considered practical  (Ref. B-38 and B-39).  Removal


efficiencies of at least 99% are expected  for such systems, although an


efficiency of 99.9% has been claimed by some vendors (Ref. 12).  Other


effluent control systems to be incorporated in these plants will include


iodine removal systems, which are expected to retain more than 99.9% of


the iodine-129 and iodine-131, and particulate filters which should

                                                                       3
reduce releases of the actinides (particulate forms) by a factor of 10


greater than the decontamination which is  effected by the chemical


separation process.  While the operating experience at NFS confirms the


effectiveness of particulate filters  (Ref  B-41), the anticipated


efficiency of silver zeolite which provides backup to the aqueous


scrubber iodine removal systems has not been verified in operational


situations.  The experience at HIS with a  different but modern iodine


recovery system has not been as good as orginally predicted  (Ref. B-40);


this equipment is now being replaced with  a silver zeolite system.  The


radionuclide releases used in this report  are estimates based on current


knowledge of the capability of waste control technology.  Technological


innovation in waste control systems may reduce projected releases, while


maintenance problems in aging equipment may result in increased


environmental contamination.

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RADIOLOGICAL IMPACT OF A REPESENTATIVE PLANT

    The assessment of the local, regional, or worldwide radiological

health Impact that may result from operation of a particular^ source or

distribution of sources of radioactive pollutants is dependent upon a

number of assumptions.  The calculational models used are generally

categorized as follows:

    1.  Source term models.
    2.  Environmental transport models.
    3.  Dose models
    4,  Risk models.

    Source term modeling includes assessments of pollutant generation

rates, inventories, and physical and chemical characteristics and

release rates to the hydrologic or atmospheric carrier.

    The environmental transport model permits an estimate of the media

(air, water, food chains, etc.) concentrations at a particular point in

the space-time continuum.  These models considered all important

pathways from the source to the receiver.

    The dose model  allows  an  estimate  of  the  energy deposition  and biological

effectiveness in a  biological system that results  from exposure to the media

concentrations.  It considers such variables  as  ingestion rates,  in-vivo

distributions and biological  half-life, and energy deposition in critical

organs.

    The risk model  provides for estimation of biological effect

resulting from doses due to ionizing radiation.

    A useful approach to evaluating the potential environmental impact

of fuel reprocessing plants and related control costs is to consider a

single plant which  is representative of current technology and design.

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Such a representative plant has been assumed to have a capacity .of 5


tonnes per day, or an annual capacity of 1,500 tonnes over an operating


lifetime of 40 years.  The fuel mix to be processed was assumed to


consist of equal amounts of uranium and plutonium fuel.  Of the residual


waste product and fissionable material in the fuel, only the following


specific radionuclides are considered in this analysis:  krypton-85, .


tritium, iodine-129, iodine-131, plutonium-239, and other actinides.


The release of other nuclides into the environment is anticipated to be

                        Q
less than one part in 10  and will not produce health effects comparable


to those produced by the nuclides considered here (Ref. D-8), although


the situation for ruthenium is unclear at present.


    Projected amounts of the radionuclides in spent fuel are given in


table A.2 through A.4, appendix A.  Liquid radioactive releases are


assumed to be insignificant, and all environmental releases are assumed


to be via the air pathway.  The population within 80 km of the model


plant was estimated to be 1.5 x 10   people by the year 1980 and was


projected to double during a 40-year plant operating life (see appendix


D).  Using these assumptions, annual dose rates from the various


nuclides were calculated  (appendix D) for individuals located 3


kilometers from the plant, and both annual and total population doses,


measured in person-rem, were determined for:  (1) the population within


80 km of the plant,  (2) the total United States population excluding


those persons residing within the 80 km zone, and (3) the world


population excluding the United States.  These doses are listed in table


2.  For tritium, most of  the United States population dose results from

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                                            10
                                          TABLE 2  .

                       Projections of Average Annual Population Dose
                      from a 5 Ml/Day Nuclear Fuel. Reprocessing Plant
Radionuclide
Source
Kr-85



H-3

1-129

1-131

Actinides
— Critical
Organ
Whole-body
Lung
Skin
Gonads
Whole-body
Gonads
Thyroid-
infant
Thyro id-
adult
Thyroid-
in£ant
Thyroid-
adult
Lung
Average
mreia/yr .
@ 3 km
0,38
0.75
13
0,50
3.2
3.2
1.4
.4
13
.8
1
Organ Dose from One
taan-retn/yr .
Regional
24
47
790
w
200
100
2.3
27
20
53
64
Year's Release for 1980 Startup
man-rem/yr . man-rem/yr .
U. S. World (less U.S.)
520 8,100
1,000 16,000
17,000 270,000
300 4,700
3,700 1,100
1,800 570
2 —
85
-
- -
400 _
I/  Decontamination factors used are 1.0 for Kr-85 and H-3, 1000 for the iodines
    and 10^ for the actinides.

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                                     11




nationally deposited tritium rather than from the worldwide distribution




of tritium.  However, most of the United States population dose from




krypton-85 is due to the worldwide distribution of this isotope.




    Local and regional population doses could vary considerably due to




differences in the population distributions and. .meteorological




conditions around specific plants as well as variations, in efficiency of




effluent control equipment.  However, estimated national doses are




probably conservative.  The projected world doses are probably correct




within a factor of five unless decontamination factors change by large




amounts.




    The estimated health effects (in terms of cancer induction and/or




mortality) for a single 5 MTU/day reprocessing plant operating over 40




years are given in table 3.  These values are based on the population




doses in table 2 and dose-effect conversion factors presented in




appendix C.  The dose-effect conversion factors are based on a recent




study performed by the National Academy of Sciences and use the




assumption of a linear non-threshold dose-effect relationship.  This




assumption is considered prudent for decision making.  The health




effects were projected on the basis of the total dose irreversibly




committed by environmental releases, and include the effects of




extremely long-lived radionuclides for the first 100 years.  The




individual risk calculation assumes that an individual resides 40 years




at a location 3 kilmeters from a reprocessing plant and also obtains his




food and water from this location.

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                                           12



                                   TABLE  3


                       Projections of Total  Health Impact
Estimated Induced Health Effects for 40 Years of Operation
Radionuclide Critical
Organ
Kr-85 Whole-body
Lung
Skin
Gonad^
H-3 Whole-body
Goaads^/
1-129 Thyroid-
Individual
(? 3 km
6.0xlO-6
l.SxlO-6
1.5X10""6
6.0xlO~6
5.2xlO~5
5.2xlO~5

Regional
Population
0.38
0.095
0.095
0.17
3.3
1.2
.012
United States
Population
8.3
2.1
2.0
3.6
59
22
.024
World
Population
130
32
32
57
18
6.9


, Total
140^
34^
*y
^
sol/
30 £ /
.04^
                infant
                          7.4x10

               Thyroid-                   ,Q2
                adult
   1-131       Thyroid-                   .12

                lQfant    4.5x10 -6
              • Thyroid-                   .04
                adult
                                                                                        9 /
Actinides      Lung       2.0xlO"6        0.13          0.4               -         0.5
          1.   50% Mortality
          2.   Very high Mortality
          3..  Low or Zero Mortality
          4.   Probably less than 25% Mortality
          5.   Genetic affects only.

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                                    13





    About two-thirds of the health effects are estimated to accrue to  the




world population outside the United States, but for this group, the indi-




vidual dose rates are extremely small (much less that one millirem per year).




The regional population group incurs about 25% of the total effects, but




receives larger individual doses than experienced by the world population group,




CmiULATIVE RADIOLOGICAL IMPACT .OF THE INDUSTRY




    In addition to determining the impact of individual plants, the




radiological impact of the entire United States nuclear fuel reprocessing




industry was assessed.  This total industry impact assessment considered




the total accumulation of the long-lived isotopes and the health effects




which can be attributed to them,  A determination was made of the total




environmental buildup of long-lived environmental radionuclides resulting




from operations of the United States industry through the year 2020.




These accumulations are shown in figure 1 and represent the estimated




cumulative environmental inventories of tritium, krypton-85, iodine-129,




and plutonium-239.  These inventories determine the magnitudes of doses




and of future health effects resulting from such cumulative environmental




contamination.                                                   ,




    Health effects attributable to the presence of these radioisotopes in




the environment were considered in terms of the cumulative health effects




that will be caused in the future due to release of these isotopes during




one year of operation of the entire industry, as well as for the




accumulated inventories shown in figure 1.




    Figures 2 through 5 represent the estimated health effects that will




be committed by the environmental buildup of certain long-lived




radionuclides if the industry is allowed to operate through a given year




at the plant decontamination factors typical of current design (1.0 for

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     10
                                     14
    10
          Figure  1. Cumulative environmental buildup of
                   radioisotopes released from total
                   fuel reprocessingindustry
    10'
    10
    10
CURIES
    10
    10'
    10
    10
I
                                                                  Df=1
                                                 10
                                                                       10
                                                10
                                                                         CURIES
                                                                  co
i	1.01
           1970       1980       1990       2000      2010
                                     YEAR
                                        2020

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                                         15
    400
Figure  2,   Estimated past and future health  effects
           committed by tritium releases from the
           United States fuel reprocessing industry
OS.
00
(_
CJ
                1980
   1990
2000
2010
2020
                                YEAR

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                                16
         Figure 3.  Estimated past and future health effects
                   committed by krypton-85  releases  from the
                   United States  fuel reprocessing industry
7000 -
             1980
1890      2000
     YEAR
2010
2020

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                                       17
     3.0
Figure 4.  Estimated past and future health effects
          committed by iodine-129 releases from the
          United States fuel reprocessing industry
                              FUTURE LIMITED  TO 100  YEARS
OS
CNI
1   2.0
€/>
    1,0
       1S70      1980       1990       2000
                                  YEAR
                        2010
2020

-------
                                         18
                         Figure 5.  Estimated past and  future health effects
                                   committed by  aetinide releases from the
                                   United States fuel reprocessing  industry
     30 -
                              FUTURE LIMITED TO 100 YEARS
o
3C
I—
CJ
««
C9
I— .

Z&
ea
t^>
(_
CJ
20
     10
       1970
                                                        2020

-------
                                    19





tritium and krypton-85, 103 for todine-129 and 10   for the actlnides).



For example, in figure 2 for tritium, if all tritium is released  from



fuel reprocessing plants with a decontamination factor of 1.0,  then by



the year 1990, the graph Shows that three estimated health effects will



have been committed by exposures received prior to 1990 and nine



estimated health effects will be caused by exposures beyond 1990  from



the quantity of tritium already in the environment in 1990.  By the year



2020, 105 health effects will have been committed by past exposures to



tritium and 250 estimated health effects will be caused by future



exposures to past releases.



    Future health effects for tritium and krypton-85 are estimated



assuming the complete decay of the quantities present.  For iodine-129



and the actinldes (238Pu, 239Pu, 2lf°Pu, 241Pu, 2tflAm, 2MfCm) only the



first hundred years of exposure beyond the time of interest are considered



for estimating health effects.  The calculational techniques used are




described in appendix Di



    These curves demonstrate a rapid change in the environmental  impact



of the fuel reprocessing industry, especially after the 1980's.   They



indicate that if reprocessing plants continue to release radioactive



material, especially krypton-85 and tritium, at current levels  of



emission (which are well below current regulations), an environmental



burden of radioactive material will accrue which presumably could result



in a significant number of avoidable health effects.  Iodine-129  and the



actinides have extremely long half-lives, and could impose additional

-------
                                   20





health risks for future generations if they remain in the biosphere




beyond the 100 years assumed for this analysis,




ECONOMICS OF HEALTE EFFECTS REDUCTION AHD EFFLUENT CONTROL




    The following economic considerations apply to a single




representative 5 tonne/day fuel reprocessing plant.  Industry totals, as




applicable, are obtained by multiplying by the number of projected




plants (appendix A).




    The fractional cost of fuel reprocessing in terms of total energy




cost was derived.  At present, the value of plutonium is approximately




$7-S/gram while the value of uranium, in the form of oxide, is about




$.Q2/gram.  For a typical 1,000 MW(e) nuclear power plant, costs of all




aspects of the fuel cycle from mining through waste disposal represent




about 18% of the total costs of power (Ref. D-16).  The reprocessing




portion alone represents about 7% of the total costs of the fuel, or




1.3% of the cost of power generation (Ref. D-16).  The capital




investment for the Allied-Gulf Company's Barnwell Fuel Recovery Plant  (5




tonne/day capacity for U02 and U02 + PuOa fuels) is approximately $80




million.  A plant of this size is capable of processing fuel from




approximately 40 to 50 power reactors, each of which cost $300 to 400




million.  Since the only reason for fuel reprocessing is to support




electric power generation by reactors, it is apparent that neither




reprocessing nor substantial emission control cost could significantly




alter the cost of power production.




    Operating costs for a 5 tonne/day plant have been estimated to be




about $13 million per year when operating at full capacity.  Of this,

-------
                                    21



approximately $3 million is for labor.  State and local taxes, insurance




and interim replacement can give rise to an estimated additional annual




expense of about $2.4 million.  The capital costs associated with




environmental studies, research, and construction is reported to



approach 301 of the capital costs of the plant, much of which has




already been incorporated in plants appropriate for reprocessing fuels




used in the uranium cycle.  For the plutonium cycle, the level of




investment for environmental considerations is anticipated to be about




twice that for processing of uranium-cycle fuel.  Present effluent




treatment capital costs are estimated to be approximately 6% of the




total capital costs.




    fable 4 summarizes, for a model plant, the estimated health effects




to the world's population attributable to 40 years of operation and the




total costs of waste control systems for the four isotopes considered in




detail by this analysis.  These total costs were computed assuming an




effective interest rate of 24%, a debt lifetime of 20 years and the




operating costs of the system over the projected life of 40 years.  A




7.5% discount factor was used to estimate the present worth of these




costs.  From these values, the cost effectiveness of the individual




component systems for control of emissions have been calculated.  The




health effects reduction as a function of system cost is displayed in




figure 6.  From these data, the overall cost effectiveness of combined




systems is generated and is displayed in figure 7 for health effects to




the. worldls population.

-------
Radionuelide
                                                 TABLE 4

                        Summary of Health Effects and Costs of Emission. Controls

                                for a 5 tonne/yr Fuel Reprocessing Plant
Removal System
   Total Cost
(present worth)
                            FILTER
a ^ 60% mortality
b * 63% mortality
c Probably less than 25% mortality
d *> 100% mortality
Health Effects
   Averted
Effects/$106
Krypton


Tritium
Iodine

Actinides

Cryogenic
Distillation
Cryogenic
Adsorption
Freon Msorption
Voloxidation (est)
Caustic Scrubber
Silver Zeolite
HEPA Filters
Sand Filter and
$ 11.5 million 270a
12.1 million
6.4 million
31 million 110b
3.5 million 303°
1.85 million 34
1 million 530d
2.1 million
23.5
22.3
42.2
3.5
86.6
18.7
530
260
                                                                                                               NJ

-------
                                     23

   S
NUCLEAR FUEL REPROCESSING - 40 YEAR OPERATION
       5 TONNES/DAY THROUGHPUT
 UJ
1200-
1000-
 80O-
 600-
 4OO-
 200-
        World Population Group
             at Risk
                                  Total Effects Avoided
                                     Projected  Mortalities Avoided
                       COMPONENT CONTROL SYSTEMS
                          ADDED IN ORDER OF
                       DECREASING COST EFFECTIVENESS
          COMPONENT CONTROL SYST.EMS

           P - PART1CULATE   T - TRITIUM
           I - IODINE       R - RESIDUAL
           K - KRYPTON
O        10      2O       3O       40       50
                    CONTROL SYSTEM COST
                         (106 Dollars)
   FIGURE 6. RISK REDUCTION vs CONTROL SYSTEM COST
                                                            60
                                                        7O

-------
                                          24
     550-
     5OO-
     450-
tn
UJ  —
^  Q 400-1
U^
uj  ;=
           P
            i
                        NUCLEAR FUEL REPROCESSING - 40 YEAR OPERATION
                              5 TONNES/DAY THROUGHPUT
                COMPONENT CONTROL SYSTEMS ADDED
                  P - PART1CULATE  T - TRITIUM
                  I- IODINE      R - RESIDUAL
                  K - KRYPTON
     200-
LU  
>-
1/5
      100-
       50-
        0
           ....
             *
       COMPONENT CONTROL SYSTEMS
       ADDED IN ORDER OF INDIVIDUAL
       DECREASING COST EFFECTIVENESS
            r^n
            x\t
                   T
                             World Population Group
                                    at Risk
                          Total Projected Effects Avoided
                                     Projected Mortality
                                          Avoided
                                                 ._|—/?-
          0
10
60
              20       30      4O      50
                CONTROL SYSTEM COST
                      (106 Dollars)
FIGURE 7 - CONTROL SYSTEM COST EFFECTIVENESS vs CONTROL SYSTEM COST
                                                                         7O

-------
                                    25




    Certain waste product removal systems are already incorporated in




present plant designs.  These include provisions for filtration of




particulates, including the actinides, and for reduction of iodine




emissions.  As is shown by the figures, krypton removal systems are




almost equally as cost effective as control systems for iodine which are




currently incorporated into plant designs when viewed, in terms of the




worldwide population,  Tritium removal systems are less cost effective,




on the same basis.




    Some of these conclusions may be modified as the underlying bases




are changed.  For example, if the calculations are based on mortality




rather than total health effects, the relative cost effectiveness of




iodine and krypton removal systems are interchanged.  Similarly, if the




calculation is based on regional effects rather than world effects, the




cost effectiveness of all systems except that for control of actinides




is sharply reduced.  These results are described in more detail in




appendix D.




SUMMARY AMD CONCLUSIONS




    The foregoing analysis of the potential environmental impact of the




fuel reprocessing industry, and of the feasibility of minimizing this,




has involved consideration of:




    1.  projection of nuclear power demand and associated spent fuel




    inventories expected to be reprocessed.

-------
                                    26


    2.  the present and potentially available technology for fission

    product removal including estimates of costs based on initial cost

    and operating expense,



    3.  the distribution throughout the environment of certain

    radionuclides released during normal operation and the resultant

    doses to the regional, national, and world populations, and



    4.  estimates of the statistical relationships of population

    exposure for specific radionuclides and the number of health effects

    expected to be associated with these.



The major conclusions derived from these considerations are as follows;



    1.  the fuel reprocessing industry, as an intergral part of the

    entire nuclear power industry, can be treated as a separate entity

    for purposes of evaluating its contribution to the overall release

    of radionuclides to the environment,



    2.  the quantites of certain radionuclides which may be released, if

    industry growth projections are substantially correct, are large.

    Consequently, the reprocessing step represents the point in the

    nuclear fuel cycle where the consequences of release of the long- •

    lived isotopes, such as tritium, krypton-85, iodine-129, and

    Plutonium 239 should be carefully considered,
                        C;

-------
                                27






3.  the consequences of the buildup of very long-lived radionuclides




(such as iodine-129 and plutonium-239) from the fuel reprocessing




industry can produce cumulative and irreversible environmental




levels which can be projected to cause adverse health effects on a




national and worldwide scale.  The consequences of the release of




the shorter-lived isotopes, while possibly more severe on a short-




term scale, are more easily reversed by effective control




techniques,








4.  control technology exists to reduce emissions of these materials.




During the near-term future, all the tritium and krypton-85 produced




during the fission process will be released to the environment




during the reprocessing stage.  Some control currently is applied




for iodine and plutonium releases.  Control technology to achieve




confinement of krypton is essentially developed and some reduction




for nearly all radioisotopes would appear feasible on a long-term




basis,








5.  removal of plutonium and other actinides is the most cost




effective method of health risk reduction.  Next in order of




effectiveness are systems for removal of krypton and iodine which




are about equal from a cost effectiveness standpoint.  Tritium




control technology is least cost effective at the present time, but




future developments should be pursued to alleviate this problem,

-------
                                28




6,  available data allows an estimate of the incremental additional




cost to the cost of nuclear production of electricity from the




imposition of waste control systems at fuel reprocessing facilities




to be about 0.1 percent of the total present cost to the producer of




electricity,








7.  evaluation of the total environmental impact of radioactive




effluents requires a consideration of dose commitments beyond those




delivered immediately in the site area, and, because of the long life




of those materials in the biosphere, must include the exposures to




national and world populations, beyond those delivered in the year




of release, and








8.  although the estimated health impact from the operation of a




single 5 tonne/day reprocessing facility is relatively small,  ,




extrapolation of the industry as a whole over the next 50 years of




operation indicates that long-term cumulative effects may be quite




large4

-------
                               29
                            REFERENCES
1.  Nuclear Fuel Services, Inc. Safety Analysis Report, AEC
    Docket No. 50-201, July 1962.

2.  Kelleher, W.J.,  Environmental Surveillance Around a Nuclear
    Fuel Reprocessing Installation, 1965-1967,  Radial Health
    Data Rep.,  10, August 1969,

3.  Logsdon, J.E., and J.W.N. Hickey,  Radioactive Waste
    Discharges to the Environment from a Nuclear Fuel
    Reprocessing Plant,  Radiol Health Data Rep., 12, 6,
    July 1971.

4.  Shleien, B.,  An Estimate of Radiation Doses Received by
    Individuals Living in the Vicinity of a Nuclear Fuel
    Reprocessing Plant in 1968,  BRH/NERHL 70-1, May 1970.

5.  Magno, P., et. al.,  Liquid Waste Effluents from a Nuclear
    Fuel Reprocessing Plant,  BRH/NERHL 70-2, November 1970.

6.  Cochran, J.A., et. al.,  An Investigation of Airborne
    Radioactive Effluent from an Operating Nuclear Fuel
    Reprocessing Plant,  BRH/NERHL 70-3, July 1970.

7.  Smith, D.G., et. al.,  Calibration and Initial Field Testing
    of 85 Kr Detectors for Environmental Monitoring,  BRH/NERHL
    70-4, November 1970.

8.  General Electric Company, Midwest Fuel Recovery Plant Safety
    Analysis Report, AEC Docket No, 50-332, September 1969.

9.  Allied Gulf Nuclear Services, Barnwell Nuclear Fuel Plant
    Safety Analysis Report, AEC Docket No. 50-332, September 1969
10. Rogers, W.M. - Nuclear Fuel Reprocessing Requirements in the
    Western United States, Western Interstate Nuclear Board,
    December 1971«

11, Oak Ridge National Laboratory, Siting of Fuel Reprocessing
    Plants and Waste Management Facilities, ORNL-4451, July 1970

-------

-------
                APPENDIX  A
SPENT NUCLEAR FUEL RADIOACTIVITY FORECASTS

-------

-------
                                    A-l
I.     INTRODUCTION
           As a starting point In the assessment of the radiological
       impact of nuclear fuel reprocessing plants on the general
       population, an estimate of the total quantity of radioactive
       materials present in spent fuels produced by nuclear electric
       power generation must be obtained.
           These estimates are based primarily on the projected
       electric power demand and on the fraction of that demand
       expected to be satisfied by nuclear plants.  To some extent,
       they are also contingent on the projected number of various
       reactor types resulting in differing amounts of the various
       radionuclides produced.  The annual reprocessing radioactive
       waste and product inventories for specific nuclides are
       forecast through the year 2020.
II.    SIZE AND SCOPE OF THE U.S. INDUSTRY
           Electrical power demand forecasts for the United States, and
       the fractional amount expected to be produced by the nuclear
       industry through the year 2020 as used in this study, are
       summarized in table A.I (Ref. A-l and A-2),
           The isotopic composition of nuclear fuels is expected to
       change over the next several decades.  Presently uranium-235
       is the most widely used reactor fuel material.  However, the
       expected increase in energy demand, coupled with the inefficient
       utilization of the fissionable material by reactors as currently
       designed, would deplete the available low-cost natural uranium
       resources, which contain only about 0.7% uranium-235, by the

-------
        Nuclear electric
Year    generation:GW(e)
                TABLE A.I

Estimated U.S. Fuel Reprocessing Requirements

    (Adapted from References A-l and A-2)

                                       *
    • Tonnes of fuel discharged annually
.Number of 5 tonne/day
reprocessing plants
required

1970
1975 - '
1980
1985
1990
1995
2000
2005
2010
2015
2020

2.6
. . 40
110
220
420
650
1000
1360
1780
2220
2700
LWR-U
25
700
1900
2700.
3700
4100
3700
3700,
4300
5300
6100
LWR-PU
0
90
500
2600
3800
4100
.- 3800
3700
4400
5400
,. 6100
LMFBR
0
0
1
0
480
2,600
'll.SOO
20,600
32,800
43,800
58,000
HTGR
0
0
1
100
2,420
6,600
. 7,800
10,000
10,000
10,000
8,800
TOTAL
25
790
2,400
5,400'
10,400
17,400
26,800
38,000
51,500 .
64,500
79,000

1
' 1
, ' 2
4
E
12
18
26
35
43
53
       Burnup:  33 GWd(t)/tonne and 0.35 thermal efficiency.

-------
                            A-3





end of the century.  Other fissionable materials, plutonium-




239 and -241 and uranium-233, are then increasingly expected to be




used to meet long-term power requirements since these can be




produced as byproducts of reactors generating electric power;




i.e., plutonium-239, -241 from uranium-238 and uranium-233 from




thorium-232.




    The types of reactors are also expected to be altered as




technology advances.  Currently, in the United States, most




nuclear power plants use light-water-cooled reactors  (LWR) of




two types, pressurized water reactors  (PWR) and boiling water




reactors (BWR),  These are fueled with natural uranium slightly




enriched to give an isotopic composition of approximately




3% uranium-235 and 97% uranium-238.  In the future, many of




these LWR systems are expected to be partially fueled




with recycled plutonium instead of uranium-235.




    Light water reactors are inefficient producers of plutonium;




the ratio of the amount of fissionable plutonium produced to




the amount of fuel consumed  (the conversion ratio) is about 1:3.




•More efficient reactor types, namely converter reactors which




produce nearly as much fuel as is used, and breeder reactors,




which produce more fuel than is used,  are expected to become




a significant part of the nuclear power industry after the




year 1985.  Examples of these two reactor types most  likely




to be used are the high temperature gas-cooled converter




reactor  (HTGR) fueled with uranium-233 and uranium-235, and the




liquid metal fast breeder reactor  (LMFBR) fueled with.




plutonium.

-------
                              A-4
    Knowledge of the relative numbers of each type of reactor



is important primarily for determination of the total quantities



of the actinides produced per unit of electric power generation.



Fission yields for most of the other elements of interest remain



nearly constant for all fissionable materials of interest.  For



this study, a mix of reactor types as given in a study by, the



AEC (Ref. A-l) was used.  Power capacity values were converted



to power generation values by using a 64% load factor



(percent of maximum available power utilized).  These were



then converted to tonnes of fuel discharged in any given year



by using data on power generated two years earlier, a thermal



efficiency of 0.35, and a burnup of 33 GWd per tonne of fuel:



    tonne of fuel discharged per year - (gigawatt (GW))



               .^ •>   tn f, power generated,    ,365 days,
    power capacity) x (0.64 •*•	*	rr—) x  (	J~~) x
            r               power capacity        year

    JL	_      Thermal power (t) •>   ,1  tonne fuel   ,

    XJ.35 electrical power (e); x (33   GW days (t)''






    A sutnmary of spent fuel projections by reactor type and



total amounts is given in table A.I.  The number of fuel



reprocessing plants necessary to service the nuclear power



industry was estimated by using these spent fuel projections



and the assumption that each fuel reprocessing plant will



handle 5 tonne of spent fuel per day (equivalent to



1500 tonne/year).  These numbers were also summarized in



table A.I for 5-year increments.

-------
                                   A-5
           The uncertainty in the above total nuclear electric




       power generation values and total spent fuel discharge values




       after 1980 is estimated to be less than a factor of two.   The




       distribution by reactor types may have a somewhat larger




       uncertainty.




III.    QUANTITIES OF RADIONUCLIDES IN SPENT REACTOR FUEL




           There are three types of radioactive material present




       in spent reactor fuel:   fission products, activation products,




       and actinide isotopes.   The quantities of specific radionuclides




       present are primarily determined by fuel type, amount of




       burnup, and time of cooling (time between removal from the




       reactor and reprocessing).




           Tables A. 2 and A.3 show quantities of the potentially




       significant fission product and activation radionuclides  present




       in one tonne of spent fuel with 33 GW(t)-days burnup and  150 days




       cooling time.  These values (adapted from Ref. A-2) are considered




       reasonably representative of all nuclear fuel types.  There is




       indication that cooling times shorter than 150 days may be used for




       some fuel cycles in the future, since quicker recycling of the




       recovered fuel produces an economic benefit.  This would




       significantly increase the amounts of shorter-lived radionuclides




       in the fuel and available for release, but would not affect the




       long-lived fission product inventories.




           The amounts of actinides estiniated to be present in




       uranium fuels and plutonium-recycle fuels are given in

-------
                                             TABLE  A.2

    Representative Quantities of Potentially Significant Fission Products  In Spent  Reactor Fuels

                                 (Adapted  from Reference A-2)
Isotope

H -3
Kr -85
fc -99
Ru -103
106
Te -125m
127m
129m
I -129
131
Cs -134
135
137
Sr -89
90
Y -91
Zr -93
-95
Nb -95
Sb -125
Ce -141
144
Pm -147
Eu -155
Half-life
(Years)
12.3
10.7
2. 13x10 5
0.11
1.01
0.16
0.30
0.09
17x10 6
0.02
2.05
3x10 6
30.2
0.14
28.9
0.16
0.95xl06
0.18
0.10
2.73
0.09
0.78
2.62
5.0
Curies per
tonne
800
10,500
15
180,000
820,000
6,500
25,000
13,000
0.04
2.0
100,000
1.2
106,000
100,000
60,000
190,000
2
400,000
800,000
13,000
80,000
800,000
200,000
40,000
Grams per
tonne
0.083
27
880
5.7
240
0,36
2.7
0.42
250
0.01
77
1400
1200
3.5
430
7.8
490
19
21
12
2.8
250
220
87
Release
state
Gas
Gas
Semivolatile
Semi-volatile
Semi volatile
Semi volatile-
Semi volatile
S emi vo la t i le
Volatile
Volatile
Semi volatile
Semivo lat i le
Semi volatile
Solid
Solid
Solid
Solid
Solid
Solid
Solid
Solid
Solid
Solid
Solid
Notes

95% released as HTO

Oxide b.p. 200 °C
Tetroxide b.p. ,80°C
169Rh + l°6Rh daughters
Oxide b.p. 750 °C
Oxide b.p. 127 fe daughter
Oxide b.p.1 29 Te daughter
b.p. 184°C
b.p. 184°C
Oxide b.p. 750 °C
Oxide b.p.
Oxide b.p. 75Q°C137™Ba daughter

90Y daughter


95ffiNb + 95ub daughters



l«rtpr + IWm daughters


Burnup = 33 GWd(t)/tonne
Cooling Time = 150 days

-------
                                      A-7





                                TABLE  A. 3




                 Representative Quantities of Potentially




          Significant Activation Products in Spent Reactor Fuels




                       (Adapted from Reference A-2)
Isotope
Mn-54
Fe-55
Fe-59
Co-58
Co-60
Half-life
(years)
0.86
2.7
0.12
0.20
5.26
Curies per
tonne
30,000
20,000
500
30,000
2.000
Grams per
tonne
3.9
8.3
.01
1.0
1.8
Release
stage
Solid
Solid
Solid
Solid
Solid
Based on 33 GWd(t) burnup/tonne




         150 days cooling time

-------
                               A-8



table A,4 (Ref. A-2 to A-8).   It was assumed that all fuels



(including those used in HTGR's) other than uranium-235 fuels,



can be considered equivalent to Plutonium-recycle fuels.



This introduces an uncertainty in these values proportional



to the deviation from this assumption.



    Based on the amounts of spent fuel to be processed, and



on the estimated quantities of radionuclides per tonne of



spent fuel, the projected annual quantities of several of



the most significant radionuclides in fuel to be processed



were calculated and are presented in table A. 5,



    Based upon plant decontamination factors appropriate for



control systems presently incorporated in fuel reprocessing


               Q     0
facilities  [10 , 10 , 1 and 1 for airborne pollutants  (actinides,



iodines, krypton, and tritium, respectively)], the annual inventories



shown in table A. 5  can be used to project the buildup of long-lived



radionuclides in the environment as a result of fuel reprocessing.



The estimated buildup of tritium, krypton-85, iodine-129, and



plutonium-239 are shown in figures A.I through A.4.  Estimates of the



environmental levels for time periods beyond 2020 by extrapolation



of these curves is  inappropriate.  Detailed discussion of the



status  of system decontamination factors is presented in



appendix B.

-------
                                       A-9





                               TABLE  A. 4




   Representative Quantities of Actinides Present in Spent Reactor Fuels




                  (Adapted from References A-2 to A-8)
Isotope
U-235
236
238
Np-237
Pu-238
239
240
241
242
Am-241
243
Cm-242
244
TOTAL
(excluding
Half -life
(years)
710x10 6
24xlQ6
4 5 10x10 8
2x10 6
86
24,400
6,580
13
379,000
458
7,800
0.45
17.6

uranium)
Uranium fuels
Ci/tonne
<1
<1
<1
<1
4,000
500
650
150,000
2
750
20
35,000
2,000
193,000

g/ tonne
8,000
4,000
950,000
600
230
8,100
2,900
1,300
510
230
100
10
25
14,000

Pu-recyele
Ci/tonne
<1
<1
<1
<1
6,000
750
1,000
300,000
5
2,000
200
250,000
25,000
585,000

fuel
g/ tonne
3,000
1,500
950,000
200
340
12,000
4,400
2,600
1,300
620
1,000
75
300
23,000

Cooling time = 150 days




Burnup =33 GWd(t)/tonne

-------
                                            TABLE  A,5

                                                                                      £
              Estimated Annual Inventories of Selected Nuclides in Spent Reactor Fuels
                                            (Curies)
Year
1970
1975
1980
1985
1990
1995
2000
2005
2010
2015
2020
Fuel Discharge
(IfC)
25
790
2,400
5,400
10,400
17,400
26,800
38,000
51,500
64,500
79,000
Tritium
2.0x10*
6.3xl05
1.9xl06
4. 3x10 6
8. 3x10 6
1.4xl07
2. 1x10 7
3. 0x10 7
41. 1x10 7
5. 2x10 7
6. 3x10 7
Krypton-85
2. 6x10 5
8. 3x10 6
2. 5x10 7
5. 7x10 7
1.1x10 8
l.SxlO8
2. 8x10 8
4. 0x10 8
5.4x10 8
6.8x10 8
8. 3x10 8
Iodine-129
1.0
3.2x10 l
9. 6x10 l
2.2X102-
4.2xl02
J.OxlO2
l.lxlO3
l.SxlO3
2. 1x10 3
2. 6x10 3
3.2xl03
Plutonium-239
1.9x10*
5. 9x10 5
l.SxlO6 -
4.2xl06
7.8xl06
1.3xl07
2. 0x10 7
2. 9x10 7
3. 9x10 7
4.8xl07
5.9xl07
Plutonium-241
7.5xl06
2. 4x10 8
' 7. 2x10 8
1.6xl09
3.1xl09
5.2xl09
i
S.OxlO9 o
l.lxlO10
l.SxlO10
1.9xl010
2.3xl01Q
Based on plutonium-recycle fuel and reactor type distribution in table A.I
(33 GWd(t)/tonne burnup and 150 days cooling period.)

-------
                                   A-ll
   700
   600
   500
   400
D
U
O
   300
   200
   100
     0
        1970
1980
1990        2000

     Year
2010
2020
        Figure A.I Estimated Cumulative Environmental Build-up of Tritium

        from the Fuel Reprocessing Industry in the United States using
        a Decontamination Factor of 1.

-------
                                     A-12
   8000
   7000
   6000
   5000
3

O  4000
D)

-------
                                A-13
0
   1970        1980         1990        2000        2010        2020
                                 Year
   Figure A.3 Estimated Cumulative Environmental  Build-up of lodine-129
   from the Fuel Reprocessing  Industry in the United States using a
   Decontamination Factor of  1000.

-------
                                   A-14
    1.2
    1.0
    0.8
•:  0.6
 3
u
    0.4
   0.2
     0
        1970
1980
2010
2020
                          1990       2000

                               Year

Figure A.4 Estimated Cumlulative Environmental Build-up of Plutonium-239

from the Fuel Reprocessing Industry in the United States using a

Decontamination Faptor of 109.

-------
                                    A-15
A-l.  Hofmann, P. L., U.S. Civilian Nuclear Power Cost-Benefit Analysis.
      Fourth United National International Conference on the Peaceful Uses
      of Atomic Energy, Geneva, Switzerland, 6-16 September 1971, A/CONF.
      49/P/072.

A-2.  OBNL Siting of Fuel Reprocessing Plants and Waste Management
      Facilities.  National Technical Information Service.   ORNL-4451
      UC-70.  (July, 1970).

A-3.  Deonigi, D. E.  Formation of Transuranium Isotopes in Power
      Reactors  National Technical Information Service.  BNWL-140 Rev.-l
      (Jan. 1966).

A-4.  Deonigi, D. E., McKee, R. W. and Haffner. Isotope Production
      and Availability from Power Reactors.  National Technical
      Information Service.  BNWL-716, Uc-23 (July, 1968).

A-5.  Burch, W. D., Bigelow, J, E. and King, L. J.  Transuranium Processing
      Plant Semiannual Report of Production, Status and Plans for Period
      Ending June 30, 1971.  National Technical Information Service.
      OML-4718, UC-80. pp. 29-30.  (Dec. 1971).

A-6.  Drumheller, K. Pacific Northwest Laboratory Division of Isotope
      Development Programs Quarterly Report Nov. 1968 to Jan. 1969.   National
      Technical Information Service,  BNWL-1010 (Feb. 1969).

A-7.  Nodvil, R. J. Supplementary Report on Evaluation of Mass Spectrometric
      and Radiochemical Analyses of Yankee Core I Fuel, Including Isotopes
      of Elements Thorium through Curium.  National Technical Information
      Service WCAP-6086'.  (Aug. 1969) .

A-8.  Crandall, J. L. Tons of Curium and Pounds of Californium
      Presented at American Nuclear Society International Meeting,
      Washington, D.C.  (Nov. 10-15, 1968).

-------

-------
        APPEHBIX   B
FUEL REPROCESSING FACILITIES

-------

-------
                                          B-l




 I.   INTRODUCTION




         This appendix provides a basic description of the processes used




     at a fuel reprocessing plant and the control systems for reducing the




     radioactive discharges.   There are three commercial nuclear fuel




     reprocessing plants capable of operation or under construction in the




     United States.   Table B.I summarizes general information for these




     three facilities.  Currently, these plants are not processing




     irradiated fuel.  The first commerical plant was Nuclear Fuel




     Services, Inc.  (NFS) which is shut down for expansion.  The Midwest




     plant is processing unirradiated fuel, and operation with spent fuel




     is anticipated for the summer of 1973.  The Barnwell plant is under




     construction, and operations are expected to begin in 1974 or 1975.




II.   GENERAL DESCRIPTION OF PROCESS




         Fuel reprocessing plants are essentially complex chemical plants,




     the complexity being compounded by the fact that the materials being




     processed are highly radioactive.  The specific process used to




     separate the spent fuel element into the product streams and waste




     stream is dependent upon the particular type of reactor fuel being




     serviced.




         All three present facilities use a shear (chop) and nitric acid




     leach method to separate the light water reactor (LWR) spent fuel




     from the metal cladding.  Following this step, the Purex process wich




     tributyl phosphate  (TBP) as the solvent is used to extract the




     uranium and plutonium from the fission product waste in column




     contactors (Ref. B-l, B-2, B-3, and B-4).  The uranium and plutonium




     are separated and further purified by various means, including ion

-------
                                              TABLE B.I

                               General Information for Commercial U.S. Nuclear Fuel
                                          Reprocessing Plants
Plant
Nuclear Fuel
Services, Inc.
Location
West Valley,
New York
Owner
Getty Oil Co.
Capacity
(MTUa/year)

300
expanding to
600 to 900
Type of Process
Purex
Midwest Fuel
Recovery Plant

Barnwell Nuclear
Fuel Plant
Morris,
Illinois

Barnwell,
South Carolina
General              300
Electric Co.

Allied-Gulf          1500
Nuclear Services,
Inc.
                 Purex
                 Aquafluor  process

                 Purex
                            w
                            NJ
 MTU = metric tons uranium

-------
                                       B-3





     exchange,  scrubbing,  evaporation,  etc.   The design recovery rate is




     99.5% for  uranium and plutonium.




         Plants designed for processing of spent fuel elements from LWR




     systems could be used to process  elements from LMFBR as long as the




     facilities are derated (handle smaller quantities of fuel) to avoid




     eritieality problems and not exceed constraints on effluents (Ref. B-




     14).




         The processing of HTGR spent  elements requires a different "head




     end" processing system since such fuel elements may require the




     burning of the graphite which contains the coated fuel particles.




     The decontamination of the off-gas stream resulting from the burning




     operation  requires development of "head end" processes unique to the




     HTGR processing facility (Ref. B-15).  The Thorex process (Ref. B-3)




     will be used for the separation of uranium and thorium from the




     fission product wastes in spent HTGR fuels (Ref. B-15).




III. PROCESS FLOW



         A simplified summary of  the aqueous  processing steps  in  plants




     under construction or in operation is displayed in figure B.I,




     The  main process steps will  be summarized here to indicate sources




     and  handling of radioactive  waste  liquids and  gases.




         1.  Cask. Unloading and Decontamination — Cask wash water is




     discharged to low-level waste system.




         2.  Fuel Pool Storage — Special provisions are made for leaking




     fuel elements storage and resulting contaminated water and off gases.




     Prevention of criticality is a major design factor.  Inventory




     control and accountability are important operating parameters.

-------
    Figure B.I   Typical process flow schematic
     Midwest Nuclear Fuel Reprocessing Plant

SOI
BBC
L

SHEAR .
V
TFAOff 1
	 1
i
f rti . . SOLVENT EXTRACTION
| 1 ACID /
1 7 Y' RECYCLE »l
PRODUCT1 RWOVTBRY >r«r«nn-t ^yy»-iyrn«Ti riT-i n rn-r/v»T
j.iiu.uuuA Miwjvjuni . WASTE CONCENTRATION 	
(SOLVENT RECOVERY) ' (ACID RECOVERY)
4-
tynnnrTnm 	 _J T
101 EXCMIGE . HIGH LEVEL
A'

PRODUCT ' 	 j — 	
^
II

i
I
I
— »-, ,-
1
i
PAHPICULATE
FILTERS
f 4^
Ag Z FILTER 1
	 	 „„ 1
1 ^
1 1 1



OFF GAS
SCRUBBER 1
1
i
I
tJ
1
pT,nT1TTrql . URANIUM SEVERAL
TnOTTr>T< FLUORIHATIOH INPUTS i

- 	 y
v
141 Y
UF^-
Lotdcut, 	 ta OT ^V11
1 	 *" WASTE
STORAGE

^
LOW- LEVEL
^ 	 WARW
CLADDING COHC1ITRATION
STORAGE . 1
• 'Lt
. Con
"J3~"" 0:
s
3W Level
:entrator
ff-Gas
irubber
1
i
i
i
i
i
i
	 j
PROCESS STREAMS

WASTE GAS STREAMS

-------
                                   B-5





    3,   FuelTransfer and Mechanical, Processing ~ NFS and Midwest




disassemble fuel elements before shearing the individual fuel rods.




Barnwell will shear the entire element and thus is essentially




limited to reprocessing fuel from light water power reactors because




of the geometries involved.  Lengths of sheared fuel rods range from




about 1.25 to 7.5 cm (1/2 to 3 inches).  Very little of the krypton-




85 and tritium is released during the shearing process.  Goode  (Ref.




B-5) reported that less than 1% of the krypton-85 is released during




this step and Cochran et al. (Ref. B-6) confirmed this conclusion




based on field studies at an operating plant.




    4.  Fuel Dissolution — In this step the spent fuel is leached




from the sheared cladding in nitric acid as a preparatory step  to the




chemical separation processes.  The leached hulls are analyzed  for




plutonium content and returned for further leaching if necessary.




Over 99% of the krypton-85  (Ref. B-5, B-6) and about 6% (Ref. B-7) of




the tritium is released into the off-gas system by this operation.




In addition to the noble gases, a large fraction of the halogens are




also released to the off gas in this step.  Most of the radionuclide




particulates present in the off gas result from this step  (Ref. B-8),




although this may not be the case at Midwest where large quantities




of particulates will be added from the high-level waste




solidification process.  Since NFS dissolves fuel on a batch basis,




essentially all the krypton-85 in the batch is discharged within a 3-




to 4-hour period.  Present NFS Technical Specifications limit fuel




dissolution to 2 metric tons per day.  Midwest and Barnwell will




dissolve on a semieontinuous basis.  Waste gases are processed

-------
                                  B-6





separately through the dissolver off gas system (DOG) at NFS and




Barnwell, since most of the radionuelides in the waste gases are




generated in this step at these facilities.  Midwest has designed




their gaseous waste treatment system somewhat differently and waste




gases are not segregated by source.




   - 5,  Chemical Separation and Purification — The nitric acid feed




from the dissolver which contains the uranium, plutonium, and fission




products is counterflowed in a contactor column with TBP and nitric




acid.  The uranium and plutonium are preferentially dissolved in the




TBP, and the fission products are retained in the nitric acid.




Separation of the plutonium and uranium in the organic solvent is




achieved by reducing the plutonium to its trivalent state where it




can be stripped with a nitric acid scrubbing process in a contactor




column.  The uranium is subsequently also stripped from the TBP which




is then recycled.  The plutonium  and uranium are then purified




through a series of processes.  Midwest has designed a calcining




system for conversion of the uranium to UF  for shipment directly to




an enrichment facility.  NFS and  Barnwell will ship uranyl nitrate in




tank trucks to conversion facilities.  Plutonium will be stored and




shipped in the nitrate form in critically safe containers.




    6.  Recovery of Solvent and Acid  — The TBP solvent and the acid




are recovered and recycled.  The  fission products are concentrated in




the evaporator bottoms of the high-level waste concentrator system




and the acid overheads are recovered.  Additional fractions of the




semivolatiles and the halogens are discharged to the waste gas system




in this step, along with some particulates.

-------
                                  B-7
    7.  High Level Waste -- At MFS and Barnwell high-level liquid




waste will be stored in stainless steel tanks which are located




underground and externally cooled.  These tanks are placed on




concrete saucers to permit monitoring for leaks and inside concrete




vaults which provide a secondary containment.  These acidic wastes




are not neutralized prior to storage to simplify future




solidification.  At the Midwest facility, which uses a fluidized bed




calcination process for solidification, high-level wastes are




solidified immediately after separation.  This process has been used




on a production scale with intermediate level waste since 1963 at the




Idaho Chemical Processing Plant (ICPP)  (lef. B-9).  The solidified




waste will be sealed and stored in a cooling pool.  The off gases




from this process will contain some radioactive particulates  (Ref. B-




8, B-9), especially the semivolatile fission products such as




ruthenium.




    8.  Low Level Waste — The low-level liquid waste stream  consists




of wastes collected from sources throughout the plant.  Midwest and




Barnwell have been designed to completely eliminate the discharge of




low-level liquid waste by adding an evaporator to the system  to




process the final low-level liquid waste stream (it is discharged to




the environment at IIS).  The evaporator overheads are discharged




through the stack and theoretically contain all remaining tritium




from the fuel.  The bottoms are solidified and are currently  shipped




to privately operated waste burial areas on site.




    In the gaseous stream, almost all of the krypton-85, and  varying




fractions of tritium and other volatile nuclides, such as iodine, are

-------
                                       B-8




     released  to  the dissolver  off gas  (DOG)  during the dissolution




     process.   Treatment  of this  waste  stream is complicated since it




     contains  varying concentrations of the  oxides of nitrogen.   NFS and




     Barnwell  are designed to have a separate waste system for the DOG.




         The various process tanks have vents for off gases which are




     contaminated with volatile fission products.  These are routed to the




     vessel off-gas system (VOG).  Midwest will combine the DOG and VOG




     systems since they have additional sources of airborne wastes from




     the calcining of the high  level liquid  waste.




         The gaseous waste treatment system  at all three facilities




     basically consists of a caustic scrubber, followed by a silver




     zeolite absorber and then  final filtration through a high efficiency




     particulate  air (HEPA) filter.  NFS has incorporated both an acid




     scrubber  and a caustic scrubber in their system.  The scrubbers and




     silver zeolite adsorption  systems  are installed to collect the iodine




     in the off-gas stream.  Midwest and Barnwell have installed or plan




     to install in series two independent particulate filtration systems




     which are isolated to avoid dual failure.  Two HEPA filter syterns are




     to be installed at Barnwell.  Midwest has installed a HEPA filter




     system in the off-gas stream and a sand bed filter system for final




     filtration of both the off-gas stream and the plant ventilation air.




IV.   CONTROL POINTS FOR EFFLUENTS




         The cladding on the fuel normally provides the primary barrier




     (or containment) for preventing the release of fission product




     wastes.   This barrier must necessarily be destroyed in the nuclear




     fuel reprocessing plant in order to recover the fissile and fertile

-------
                                  B-9





material for reuse.  This fact, in addition to the very large




quantities of fission products present, requires that effluent




control procedures must be incorporated at all processing steps.  In




practice, control of the effluents from each process can be achieved




through the use of a common collecting system such as the vessel off-




gas header system which collects the off gases from several




processing vessels.  The use of such common systems has definite




economic advantage.  Fewer control systems are required and a higher




degree of reliability for control of discharges is possible since




there are fewer components subject to failure, better quality




equipment can be installed.




    An example of a common collection system is the low-level liquid




radioactive waste system.  Sources feeding this system include, but




are not limited to: cask decontamination water, leakage from the fuel




storage pool and waste storage tanks, laboratory wastes, laundry




wastes, high-level waste equipment drains, and floor drains.  This




waste is collected and processed through an evaporator where the




overheads can be condensed and recycled, discharged, stored, or




handled in a combination of these options.  NFS, Inc. has chosen the




method of discharge of the condensed overheads through a system of




settling lagoons to a public waterway.  Midwest has chosen a




combination of treatments where the evaporator overheads are




condensed and recycled to the maximum extent possible.  Low-level




liquid wastes which cannot be recycled, such as air scrubbing wastes,




are collected in a low-level waste vault which is maintained at a




constant volume by use of a second evaporator.  The overheads of this

-------
                                B-10




evaporator are not condensed but discharged through a second air




cleaning system and the sand filter to the stack.




    In general, air cleaning systems follow the same principle.




Contaminated air is collected from various processes in a common




header and then treated to remove contaminants.  Reprocessing plants




have sufficient chemical contaminants in some off-gas streams to




cause problems, such as overloading of various air cleaning systems.




Therefore, the off-gas streams are frequently segregated by source,




especially for initial treatments.




    Specific effluent control points and the principal contaminants




are; (1) Dissolver Off Gas — noble gases (krypton-85 and xenon-133),




halogens (iodine-129 and iodine-131), tritium, and particulates.   (2)




Solvent Extraction and Purification Off Gas — particulates.   (3)




Solvent Recovery Off Gas — mixed fission products.   (4) Acid




Recovery Off Gas (High-Level Waste Concentration) — halogens and




other volatile species.  (5) High-Level Waste Solidification —




ruthenium and other potential volatile radionuclides including




technetium, cesium, selenium, and tellurium (Ref. B-8).   (6) Low-




Level Waste Treatment — tritium, strontium, ruthenium, cesium, and




other longer-lived radionuclides  (Ref. B-10).




    Investigations are being conducted into modifying spent fuel




reprocessing systems to provide more positive control of  the




effluents.  ORNL is presently directing efforts  to the design of




equipment and process flow to obtain "near zero  release"  processing




of short-cooled LMFBR fuel (Ref.  B-ll).  The design includes a  new




"head-end" processing step (voloxidation) that is designed to release

-------
SECONDARY
GAS
TREATMENT






MJT3
FINAL
OFF-GAS
TREATMENT
mjr>TO *TTJ
                                                                    TO STACK
SPENT
FUEL
 RECEIVING
  THROUGH
VQLOXIDATION
DISSOLUTION
  THROUGH
    FEED
ADJUSTMENT .
SOLVENT EXTRACTION
 THROUGH PRODUCT
  CALCINATION
HN03- H20
RECOVERY
                                            PuO,
                                                          HNO
                                               RECYCLE
                                                             3*"1
                                                          H2°
    WASTE
SOLIDIFICATION
AND SOLID WASTE
   PACKAGING
                                                                              STORAGE
               Figure B.2  Schematic diagram showing the steps required for the
                           "zero release" reprocessing concept (B.ll)

-------
                                   B-12


    tritium from WFBR fuel and deactivate sodium prior to aqueous

    processing.   The tritium, krypton, and iodine are evolved during the

    head end operations (which include voloxidation and dissolution) and

    vented to their respective primary removal systems.  The off gas from

    the dissolution, cell and from the process equipment beyond

    dissolution and feed adjustment is subjected to a secondary off-gas

    treatment system which includes filters for partieulates and

    scrubbers for the oxides of nitrogen, halogens, and ruthenium,

    Argonne National Laboratory is also developing an alternate "head-

    end" pyrochemical process for decladding of LMFBR fuel which

    potentially may produce an improved method for control of the

    effluents (Ref. B-12, B-13).

V.  SOLID WASTES

        The solid wastes resulting from the recovery of uranium and

   , plutonium can be categorized as high-level solidified wastes, spent

    fuel cladding wastes, and low and intermediate level wastes.  Federal

    regulations (Ref. B-16) require that high-level wastes, generally

    interpreted as self-heating, be solidified within five years of
           *
    processing.   In addition to the radiation exposure protection which

    must be provided for these wastes, cooling must also be provided.

    The spent fuel cladding wastes contain residual amounts of the fuel

    in addition to the activated metallic radionuelides making up the

    cladding itself.  There are many sources of low and intermediate

    level wastes: air filters, spent resins, silver zeolite, evaporator

    bottoms, sand filters, etc.  It appears the greatest problem

    presented by these wastes is the presence of long-lived radionuelides

-------
                                    B-13





    of health significance such as the alpha-emitting transuranics and




    lodine-129.   These long-lived components will be present in all solid




    wastes from reprocessing plants,




VI. DISCHARGE CONTROL OPTIOMS




        In general, the fuel reprocessing industry has incorporated the




    most advanced technology into their waste treatment systems.  For




    example, the control system for iodine, one of the limiting




    radiomiclides in the local environs of a reprocessing plant,  (Eef, B-




    17) will use silver zeolite technology which has only recently been




    developed.  It should be noted that a control method is not available




    for tritium and only the one control system is planned for iodine.




        Most of the radioactive discharges to the environment from




    reprocessing plants will be in the gaseous waste Affluent.  The two




    newer plants have designed their processing systems to eliminate the




    discharge of liquid radioactive material.  However, neither of these




    facilities has been operated with irradiated fuel.  Thus, a decision




    on whether it is preferable to discharge radioactive waste to the




    atmosphere only or to discharge via both the liquid and gaseous




    pathways should be postponed until operating experience with both




    methods is obtained.  Since most of the current efforts to reduce




    discharges have been directed toward gaseous effluents, the methods




    discussed in this section are limited to the control of airborne




    discharges.  However, liquid waste discharges may require additional




    investigation in the future.




        The radioactive pollutants that are most likely to be released




    from normally operating reprocessing plants are krypton, tritium,

-------
                                B-14





 iodine,  and the actinides.   Other fission products  such as strontium,




 ruthenium,  and cesium,  and  induced activities in the fuel element




 cladding can also be released.   In addition to the  anticipated normal




 discharges  (gaseous waste stream), miscellaneous airborne releases




 can occur because of the complexity of the various  processing oper-




 ations  and  the unproven reliability of some of the  control systems.




 Such releases may not be detected by monitoring of  the gaseous waste




 stream  (stack effluent). Inplant air monitoring for contamination




 control can, however, indicate  these possible pollutants.  A summary




 of  the  characteristics  of gaseous waste control systems for the




 isotopes of major concern is presented in table B.2.  It is seen that




 the long-term operational reliability of control systems is unproven




 with the exception of the H1PA  (high efficiency particulate air) and




 sand filter systems.




.    KRYPTON-85 CONTROL




     Up  to the present,  krypton-85 and other noble gases have been




 released directly to the atmosphere from nuclear reactors and fuel




 reprocessing plants. With  no off-gas treatment for noble gases about




 10,000  Ci of krypton-85 is  estimated to be discharged per metric ton




 of  spent fuel processed, assuming a burnup of 33,000 megawatt days




 per metric  ton.




     Several methods have been suggested to limit such releases.  The




 processes are classified as ambient temperature adsorption, cryogenic




 adsorption, cryogenic distillation, selective absorption,




 permaselective membranes, and clathrate precipitation.

-------
                                               TABLE B.2
                           Control System Data for Nuclear Fuel Reprocessing
                                         (LWR + Recycle Fuels)
                                                  System
                                                  Decon
  Projected
COSTS
Isotope

Krypton-85





Tritium


Iodine-131


Iodine-129


Aetinides




Control System

(a) None
(b) Cryogenic
distillation
(c) Cryogenic adsorp-
tion (charcoal)
(d) Freon adsorption
(a) None available
(b) Voloxidation

(a) None
(t>) Scrubber
+ AgZ
(a) None
0>) Scrubber
+ AgZ
(a) None
(b) Pre~filter
+ 2 HEPA's
(c) Pre-filter + HEPA
+ Sand Filter
Reliability
D
NA

Good

Unproven
Unproven
NA
Unproven

NA

Unproven
NA

Unproven
NA

Good

Excellent
Factor

1
a
io3
2
102
10
1 2
10

1

IO3
1

10 3
io5-io6

io9

io9
Release Rate

10 Ci/MTU

10 Ci/MTU
2
10 Ci/MTU
10 ' Ci/MTU
800 Ci/MTU
8 Ci/MTU

2.0 Ci/MTU

0.02 Ci/MTU
0.04 Ci/MTU

0.0004 Ci/MTU
0.6-6 Ci/MTU

6xlO~ Ci/MTU

6x10" Ci/MTU
Capital/plant

NA
c
$3x10
6
$3x10
$1.5x10
NA
Unknown

NA
c
$1.2x10
NA. -
c
$1.2x10
NA

$1.0xl05

$3. 5x10 5
Operation/yr

NA
R
$1x10
5
$1.5x10
$1x10
NA w
Unknown 1
tn
NA

Unknown
NA

Unknown
NA

%
$5.0x10

$7.0xlo'*
Notes:
  1)  NA indicates not applicable
  2)  Iodine-131 content is estimated for fuel with 33,000
MWd/MTU burnup and 150 day cooling

-------
                                 B-16






    The processes have been previously reviewed by several authors




(lef. B-18, B-19, B-20, B-21, and B-22) with regard to development




status, advantages and disadvantages, cost, and efficiency.  Kirk




(Ref. B-18) has prepared a comprehensive review of the radiation




hazard from krypton-85..  The conclusions reached in these reviews




indicate that at present only cryogenic distillation, selective




absorption processes, and eryegenic adsorption are worthy of




consideration for control of krypton-85 discharges from reprocessing




plants.  (See table B.3 from Ref. B-19.)  Systems based upon both the




cryogenic adsorption and cryogenic distillation processes have been




designed for and are being installed at light-water-reactors for




extension of holdup times for gaseous effluents containing noble gas




radionuclides.  The selective absorption process has been developed




for application to reactor systems, but requires further development




to be applicable to fuel reprocessing plants  (Ref. B-20).




    The cost of krypton collection systems is highly dependent on the




design of the dissolution process.  Essentially all of the, krypton




present in the spent fuel is released during  this dissolving or




leaching process.  To minimize the costs, it  is necessary to minimize




the total volume of off gases from this process since all of the off




gas must be treated to remove the krypton.  Therefore, in a new plant




the cost of a krypton collection system would probably be




significantly less than installation"of such  a system in an operating




plant where no effort was'made to minimize the total off gas from the




leaching process.  The costs presented are typical of costs for a

-------
                                         TABLE  B.3
COMPARISON OF PROCESSES FOR THE REMOVAL OF 85Kr FROM DtSSOLVER OFF-GAS
     A FUEL-REPROCESSING PLAIT '
Process
1. Room -temperature
charcoal beds or
molecular sieves
2. Low-temperature
cKsrcoal beds or
silica gel
3. Cryogenic
distillation
4. Liquid extraction
5. Clathr ate precipitation
Kryptonates
6. Perm-selective
membranes
7. Thermal diffusion
8. Electrostatic
diffusion
Kr
recovery
Oft)
• 99
99
98
99
Unknown
99


Development
status
Bench scale completed;
scale -up feasible.
Development completed;
plant operated
Developed and
operated en a
significant scale
Bench scale completed;
demonstration needed
for large scale
Laboratory studies
only? no engineering
Bench-scale worki
need engineering tests
Little pertinent data
Limited; technical
feasibility not proven
Advantages
Simple operation;
accepts dilute feed
gas
Small-volume beds;
uses dilute feed
gas
Low capital cost and
low operating cost
Using Freon-12: low -
refrigeration costs;
low solvent costs;
no explosion hazard;
might eliminate
pie -treatment
ssKr is collected
as a solid;storable



Disadvantages
Large-volume adsorber beds;
charcoal can ignite; . strong
oxidizing gases must be removed
prior to adsorption
Charcoal can ignite; oxidizing
gases, CO2 and tL^O must be
removed! large consumption
of liquid nitrogen: adsorbers
must withstand high pressure;
high operating cost.
Explosion hazard in forming and
concentrating ozone
The absorber column operates at
200 lb/in2 (gauge); the volume of
extractant is large if operated
at 15 Ib/in2 (gauge)
Meeds concentrated feed gasj
crystallization step slow
Membranes sensitive to chemicals;
high power costs

Poor economics for disposal
of dilute 85Kr waste
This Table is reprinted with the kind permission of Dr. C. M. Slansky who presented it in
a paper published in Atomic Energy Review  CSee reference B-19)

-------
                                     B-18





krypton control system In a new facility.  At the Midwest facility an




effort was made to prevent large scale dilution of the off gas.




    Cryogenic adsorption systems remove krypton, .from process gas




streams by adsorption in refrigerated activated charcoal beds until




the bed capacity is reached, followed by desorption into a purging




gas while heating the beds.  This was demonstrated on a large scale




at the Idaho Chemical Processing Plant (ICPP) more than 15 years ago.




The disadvantages of this process are high refrigeration costs, fire




hazard potential, explosion potential due to hydrocarbons, nitrogen




oxides and ozone, and impurity plugging of the adsorbers.  An overall




recovery fraction of 311 was obtained for krypton although with




design changes and modifications in operating procedures, a recovery




fraction of the order of 99% could be achieved.  The system can be




considered but may not be the best for application to fuel




reprocessing plants.  It may have more potential for application to




interim holdup for effluent gases in reactors.




    Experience has been gained for this method in the development of




the HTG1 and a decontamination factor of 10  appears to be achievable




under most operating conditions.  Operation of the adsorber beds at ,




cryogenic temperatures helps overcome the problem of an occasional




abrupt release of adsorbed contaminants which has been experienced




with ambient temperature absorbed beds.  Assigning a decontamination




factor to ambient temperature systems is questionable since




experience has shown that under adverse conditions it is possible to




experience a negative decontamination factor for interim periods.

-------
                               B-19



    Total capital costs, including installation, for a cryogenic




adsorption system are estimated at 3 million dollars based on general




estimates for use of systems at reactors.  Slansky (Ref. B-19)




estimated one million dollars for capital costs which appears low.




An annual operating cost of $150,000 is reasonable.




    Cryogenic distillation, which is based upon separation of gases




due to differences in their relative volatilities at low




temperatures, has been demonstrated and operated on a significant




scale at ICPP for removal of krypton and xenon from an off-gas stream




(Ref. B-23 and B-24).  Recovery of krypton and xenon in a form




suitable for bottling in gas cylinders is possible with this process.




An additional advantage is the lower capital and operating cost of




this cryogenic system as compared to cryogenic adsorption.  The




cryogenic distillation process entails some potential for explosion.




In spite of the explosion potential, cryogenic distillation is




considered to be one of the two most promising processes for noble




gas control at reprocessing plants.




    Considerable experience has been gained in operating these




systems in liquified gas (or air) plants.  While decontamination


                                        3

factors across the cryogenic stage of 10  have been estimated, small




leaks can occur in a system.  Estimates of the recovery factor for an




overall system range from 98% to 99.99%.  Decontamination factors of



  3
10  (99.9% recovery) should be attainable with this system and




guarantees of such performance have been submitted to the General




Electric Co, with bids for installation of a cryogenic distillation




system at their Midwest plant  (Ref. B-26),

-------
                               B-20





    The General Electric Co. received bids ranging from 0.75 to 1.5




million dollars for the equipment needed to install a cryogenic




distillation system in their Midwest plant (lef. B-26).  It is




estimated that installation costs would match equipment costs, thus




producing a total capital cost of about 3 million dollars for this




system.  Annual operating costs of $100,000, postulated by Slansky




(Ref. B-19)» are probably a good estimate.  Xenon is also recovered




by this system and if kept separate from the krypton, has a potential




market value.




    Selective absorption (liquid extraction) depends upon the




relative solubilities of gases in the solvent used — Freon-12 being




the typical solvent under consideration.  Krypton and xenon are




selectively absorbed in this solvent while other materials pass




through.  The solvent is then processed to recover the krypton and




xenon which can be stored while the solvent is recycled.  Bench-scale




studies have been completed (Ref. B-20 and B-25) and, as indicated




previously, commercial systems for application to reactors have been




developed.  Because of the anticipated explosion hazard associated




with the cryogenic systems, management personnel of commercial fuel




reprocessing facilities (NFS, Barnwell) have indicated a preference




for fluorocarbon selective absorption systems of this type.




    Major disadvantages of such a system include radiation




degradation of the solvent, and a requirement for pretreatment of gas




streams to increase system tolerance to impurities.  Babcock and




Wilcox indicates that these problems can be circumvented with good

-------
                                    B-21





    engineering design.  Decontamination factors of 100 to 1000 are




    expected for such systems (Ref. B-20).




        The cost of a selective absorption system for installation at a




    reactor is estimated at one million dollars.  Assuming a slightly




    greater capacity system for a reprocessing plant results in a total




    installed cost of $1.5 million.  Annual operating costs of $100,000




    predicted by Slansky (Ref. B-19) also appear reasonable.




VIII.  ' TRITIUM.CONTROL




        Processes available for tritium control have recently been




    reviewed (Ref. B-27).  The techniques considered include: chemical




    exchange, distillation, electrolysis, diffusion and centrifugation,




    radiolysis, adsorption and chromography, solvent extraction, and




    molecular excitation.  At present the information on these techniques




    is inadequate to project the technical or economic feasibility of




    retaining tritium.




        Voloxidation is a new method of tritium control that looks




    promising (Ref. B-ll),  In this process the fuel pins, just after




    shearing, are heated to approximately 650° C in a stream of air or




    oxygen.  Tritiated water is generated which should be relatively free




    of ordinary water and consequently occupy a much smaller volume than




    tritium wastes from presently planned 6r operating reprocessing




    plants.  This process may collect approximately 99% of the tritium




    which is present in the unprocessed fuel.  Since the voloxidation




    process requires a major change in the head-end design of fuel




    reprocessing plants, it would probably be impractical to back-fit




    existing or planned facilities.

-------
                                    B-22





        The projected release rate of tritium•from a facility having no




    tritium control systems is 88 Ci/MTO.   Releases may be either by the




    air or water pathways depending on plant design.  With the




    voloxidation head-end process, tritium decontamination factors of 100




    are anticipated.  Cost information on  the voloxidation process is not




    presently available5  however, General  Electric considered three other




    methods of tritium recovery and/or disposal in connection with their




    Midwest plant (Ref. B-26).  Removal from off gas was stated to cost




    approximately $10 million with no process or technical feasibility




    defined.  Deep-well disposal of tritiated water was reported to run




    between $400,000 and $500.000.  Finally, off site shipment cost was




    estimated to be between $250,000 and $350,000 with no estimate of




    feasibility or safety associated with  the packaging and transport,




IX. RADIOIOPIN1 CONTROL




        A variety of processes have been developed and used for




    collection of radioactive iodine.  These processes include (1) wet




    collection — aqueous scrubbing (reactive sprays, towers, wet




    filters); (2) adsorption (charcoal, activated charcoal, silver




    zeolite metallic filters); and  (3) filtration (high efficiency




    particulate — HEPA,, sand deep-bed fiber glass).




        The particular treatment process to be selected depends upon the




    collection efficiency desired, quantity and chemical species of




    iodine involved, stream characteristics, i.e., flow rates,




    temperature, humidity, etc.  The actual overall removal efficiency of




    a system is dependent upon the off-gas flow paths as well as the




    characteristics of the treatment techniques.

-------
                                B-23





    WetCollection Techniques have been used to quantitatively retain




iodine in the liquid phase by adding mercury salts during fuel




dissolution at the Savannah River Plant.  The resulting Hg-I complex




is solvent extracted and the solvent is washed in a solvent scrubber




to remove any remaining iodine.  Disposal of the solid Hg-I complex




must be effected.  However, there are no AEC guidelines for disposal




of such waste at present.




    A mercuric iodate precipitate can result from scrubbing the off




gas with a mixture of 8 to 14 molar nitric acid and 0.2 to 0.4 molar




Hg(NO,,),j (mercuric nitrate).  Solid iodine can be obtained by use of




concentrated 17 to 19 molar nitric acid at room temperature.  The




latter systems can be considered capable of removing massive




concentrations of elemental iodine and trace quantities of other


                                     k

forms.  Decontamination factors of 10  for all forms of iodine are




reputed to be theoretically feasible in either packed or bubble-




packed scrubbers at operating throughput rates considered appropriate




for fuel reprocessing (Ref. B-28).  If the off gas contains high




concentrations of NO  some loss in the decontamination factor occurs.




Alkaline solutions (NaOH, NaHC03, NaC03) are also used in scrubber




columns and have reported decontamination factors of 10 to 20,  In




packed columns, the separation  efficiency is a function of bed




height, fiber drag coefficient, fiber diameter and fiber volume




fraction.




    An operational facility  (Euroehemi Fuel Reprocessing Plant) which




uses  two scrubbers (NaOH for low iodine concentrations and NaHCO  or
                                                                w


Hg(NOg) for high iodine concentrations) reports decontamination

-------
                               B-24




factors of 500 for gaseous iodine and 2,500 for iodine in aerosols.




The chemical form of iodine which emanates from these scrubbers is




predominantly organic (Ref. B-29) although traces of hypoiodous acid




have also been reported.




    Reactive sprays, hydrazine and thiosulfate, have been studied for




application to iodine and methyl iodide washout from reactor




containment atmospheres with decontamination factors of 2,000 and 100




reported respectively.  Such systems have not been applied to fuel




reprocessing plants.  Their performance is found to be a function of




relative humidity, temperature, drop size, solution pH, and




concentration,




    Charcoal Adsorption has been used for more than 10 years in




reactor and fuel reprocessing off-gas systems.  The iodine removal




efficiencies of activated charcoal that have been reported cover a




broad range (50-99.99%).  The actual decontamination factor is




dependent upon the forms of iodine, concentration, stream humidity




and flow velocity, and charcoal impregnant.




    Silver zeolite (AgN03 impregnated in an alumina-silica molecular




sieve) is reputed to be superior to impregnated charcoal for the




adsorption of methyl iodide (approximately 20 times that for




impregnated charcoal under dry air conditions)(Eef. B-30).  The




principal design consideration is the effective residence time in the




bed which is related to the face velocity.  If the silver zeolite




(AgZ) beds are designed to permit a mean residence time of about 0.5




seconds, efficiencies are greater than 99.91 under the most adverse




conditions of humidity and chemical form of the iodine (Ref. B-30 and

-------
                               B-25






B-31),  Eemoval efficiencies  can be optimized  through  selection of




the  zeolite material with  its characteristic alumina-silica ratio.




However,  this  ratio largely determines  the resistance  of  the AgZ to




acid vapors and thus the effective life of the bed.  In general, high




acid resistance results in lower removal efficiencies.




     Although the  efficiency, of the AgZ  is acceptable,  consideration




must be given  to  the loading  characteristics of AgZ  in terms of the




total iodine cleanup system.   The  AgZ beds will adsorb all halogens




and  probably cannot be used when HC1  is used for fuel  dissolution.




In addition, there is  a considerable  amount of iodine  (iodine-129 and




-127) present  in  the off gas, and  this  may load the  AgZ at




unacceptably high rates.   Current  technology employs the  alkaline




scrubber  to remove a large fraction of  this iodine which  is




theoretically  elemental in composition.  Work  is continuing on the




development of other metallic zeolites  which may eventually replace




the  scrubbers. Lead zeolite  currently  appears to be the  most




attractive of  these more economical and less efficient systems.




     The lifetime  of the AgZ system is improved by introduction of an




oxidizing catalyst upstream of the sorbent (Ref. B-28),  The use of




AgZ systems  has  been  shown to be feasible and  practicable, but full




scale operation has yet to  be  accomplished.  Adsorption of iodine on




metallic  filters  (copper,  steel or aluminum, silver-coated copper)




has  resulted in  recovery in the range of 97.4  to 99.9%.  Silver-




coated  silica  gel has  a reported decontamination factor of 10  for




iodine  (elemental).

-------
                               B-26






    Filtration is primarily designed for control of participates.  It




cannot be seriously considered as a primary technique for iodine




collection, since it depends upon sorption of iodine to particulates




which are then trapped by the filter.  In any event, all off-gas




streams from fuel reprocessing plants will be filtered through sand,




deep-bed fiber glass or high efficiency particulate air filters in




addition to the iodine control systems.  The iodine collected by




particulate filters will experience desorption at a rate which is




dependent on stream conditions.  While this desorption rate may be




significant for the short-lived iodine-131 discharge, it will have a




negligible effect on the iodine-129 discharge.  Therefore, the




removal efficiency for filtration of iodine is negligible,




    Control Systems Evaluation.  It is estimated (Ref. B-32) that




much of the discharged iodine will be in an organic form and that




most of this iodine will interact with atmospheric particles and




settle out within 10 to 20 miles of the discharge point.  Thus, the




discharged iodine appears to be principally a local problem, although




long-term environmental transport must be considered because of




iodine-129 with its long half-life.




    LWR spent fuel with a burnup of 33,000 megawatt days per metric




ton will contain about 2 Ci of iodine~131 per metric ton after 150




days cooling.  In addition to being dependent on the irradiation .




history of the spent fuel, the iodine-131 content decreases rapidly




with the cooling period because of its 8 day half-life.  The iodine-




129 content will be about 0.04 Ci per metric ton of spent fuel.

-------
                               B-27





    An overall system for iodine collection in fuel reprocessing




plants will probably consist of a combination of wet collectors




(liquids and off-gas scrubbers) followed by a catalytic decomposition




system, an adsorption system, and a filtration system.  The overall




efficiency will be dependent upon the detailed design of the off-gas




flow system.  For example, if the system is assembled in the manner




of the Midwest facility, the off-gas stream bearing the iodine is




first processed through the caustic scrubber (eff. - 90%) and then




through the AgZ bed.  The gas stream is then routed to the sand




filter and stack.  The overall removal efficiency for this treatment




should be greater than 99.9%.  However, the scrubber solution,




theoretically containing 90% of the iodine, is routed to the low-




level-waste storage vault.  The low-level waste in the vault is




routinely evaporated for volume reduction, opening a way for a large




fraction of the iodine to be revolatilized.  This second off-gas




stream is processed through.a second scrubber but no AgZ bed.




Therefore, about 10% of the iodine routed to the low-level vault can




be discharged to the sand filter and the stack.  The buildup of the




long-lived iodine-129 in the waste storage vault must also be




considered in this system since it is reasonable to assume that the




iodine-129 discharge will increase with the inventory buildup.




    The development work previously discussed may produce more




effective iodine cleaning systems for reprocessing plant off gases




(lef. B-ll, B-12, and B-13).  However, even with present technology




it is difficult to speculate what the cleaning efficiencies of actual




installed systems will be.  Once these systems are in operation their

-------
                               B-28




performance can be monitored and documented.  In the interim the use



of an overall decontamination factbr of 100 to 1,000 appears



acceptable.



    Based on experimental evidence it is reasonable to assume that



the wet scrubber/AgS system can be^designed to achieve a minimum



decontamination factor of 1,000 under most conditions,  However, as



discussed previously, the performance of the off~gas iodine cleaning



system may not be the controlling factor in determining total iodine



discharges from a reprocessing plant.  In particular, the system for



handling or processing the scrubber solutions must minimally provide



the same cleaning efficiency as the off-gas stream since a large



fraction of the iodine is expected to be in the scrubber solutions.



Therefore^ it can be concluded that while the efficiency of currently



planned iodine removal systems will be a minimum of 99,9%, the total



iodine waste handling system introduces a large uncertainty.



    Meager data are available regarding the costs for iodine removal



systems.  The costs are dependent upon the volume of off gas which



requires processing as well as the total quantity of iodine to be



removed.  The following estimates have been adjusted to 1970 dollars



using the Marshall and Stevens Equipment Index.

-------
                               B-29
Wet Collection           Capital Costs*    •••  Operating Costs*

a.  Spray tower          $310-620/1,000 cfm    Unknown
b.  Packed towers        $620-1,240/1,000 cfm  $510-1,020/1,000 cfm
c.  Solvent extraction   Unknown               Unknown          .

Absorbing Systems:        •    '         ••     '        ' '

a.  Charcoal             $40,000/bed           Unknown
b.  Charcoal + catalytic
    combiner (KRB)       $750,000^$10          Unknown
c.  Ag zeolite           $400,QQO/bed          Unknown

Complete SystemEstimate;

5 MTU/day throughput              g                     k
a.  Caustic scrubbers    $7.5 x 10             $7.5. x 10
*  At an average plant the total air volume moved is approximately
   60,000 cfm.

    The costs listed above are for the caustic Scrubber systems  only

since the quantities of AgZ needed will not be known until  operating

experience is gained to determine the extent, of  the .parasitic  loading.

-------
                               B-30


the alternative of using holdup for control of iodine-131 from fuel

reprocessing plants is applicable only in conjunction with a removal

system that will extract the long-lived radioiodine, iodine-129.

Estimates of the cost involved for holdup are as follows:

    Given:  plutonium price                        $7/g
              annual interest rate                   10%
              LWR fuel plutonium content             16* kg/MTU
              LMFBR fuel plutonium content           86 kg/MT
              time increment (1 month)               30 days

    MlES.  x II x MP. x i-2SSE    . $930/tonne month
    MOT       g    year   12 month

Example:  Barnwell will have a capacity of 1,500 MTU/year and will

process at 150 days cooling time.  If the cooling time is increased to
                                                                  6
300 days to permit the iodine-131 to decay by a factor of about 10 , the

cost will be:
    930  $        1.500 MTU             .$6.975.000
    MTU month   x   year    x 5 »°at*8 "   year

    For each additional 30 days of holdup the cost would be:

    930  $	, x 1,500-mu x l month m $1,395,000
    MTU month        year                  year
    For LMFBR fuel cooled for 150 days, the cost would be approximately

10 times greater.

*Based on 2/3 UO,,  fuel and 1/3 mixed oxide fuel.   (See  table A-4)

-------
                                   B-31
X.  PMTICUIATE CONTROL                     •
        Radioactive particulates associated with fuel reprocessing will
    be found in both the liquid and the gaseous effluent streams.  The
    particulates arise due to the various operations (shearing through
    waste solidification) and as a consequence can include a variety of
    radioisotopes.  They will also be in a variety of chemical and
    physical forms.  Easily condensable vapors should be considered as a
    source of particulates bearing radioactivity.  Most of the
    radioactive wastes available for discharges are attached to
    particulates.  Included in this category are the isotopes in the
    actinide series, many of which are highly radiotoxie since they decay
    by alpha emission.  Other radionuclides which potentially can
    contribute significantly to the waste gas stream particulate makeup
    include the volatile isotopes, such as ruthenium-106, cesium-134, and
    the longer-lived isotopes, such as strontium-90 and eesium-137.
        The particulate effluent from reprocessing plants will for the
    most part be soluble and will quite probably be in the nitrate form.
    The particulate effluents from the solidification process will
    probably be oxides and thus insoluble.  Ruthenium may be an
    exception, especially from the solidification process, and may be
    complexed as an organic.
        Probably the most acceptable theoretical estimates of discharged
    radionuclides were made by Oak Ridge scientists (Ref. B-8).  These
    estimates appear conservative, however, as indicated by the gaseous
    discharges measured at the NFS facility (Ref. 1-17).  An annual
    discharge of 1.0 to 10 Ci of beta-gamma fission products, excluding

-------
                               B-32



tritium, noble gases and the iodines, currently is the most


reasonable estimate.  The routine annual gaseous discharge of


actinides at the DSAEC Rocky Flats Plutonium Recovery Facility  (Ref.


B-37) averaged; 2i4 millicuries from 1953 to 1970.  Fuel reprocessing


plants may be expected to have higher discharges (see table B.2).


    Control Techniques - Particulars are generally collected from


gaseous waste streams through the use of inertial separators


(cyclone, or gravity settling), filtration  (fabric, glassfil,


sandbeds, HEPA), precipitation (electric, thermal) sonic


agglomeration, and liquid scrubbing.  The specific system used  is


dependent upon the type of source and efficiency of control desired.


   '.Filtration - Because of the concern regarding the inhalation of

••• . '•>.  •' '. •    '•''.''
radioactive particulate material, high efficiency filtration systems


for removal of particulates from air have received the primary

                     -.-'.'                        "       i
emphasis in the nuclear industry and are widely used  (Ref. B-32).


farticulate filters and materials of a variety of types are


available.  The materials include cellulose-asbestos, glass, glass-


asbestos , plastic fibers and ceramics.  Filters are generally


classified as panel  (viscous impingement) filters which are coated


with a tacky substance to increase particle adherence, and extended


medium dry-type filters which are called "bag" or "sock" filters.


Panel filters are designated Group I and have a low efficiency  for


small particulates.  Bag filters are designated as Group II and Group


III (medium and high efficiency, respectively).  HEPA (High


Efficiency Particulate Air) filters are classified as a special


group.  Sand filters (Ref. B-33) in general are designed for a

-------
                               B-33



specific application, usually for use in cleaning corrosive materials
                                           '.'.'  •".-('•     ' ' •   ' ' '

from air streams.  Their use in the nuclear field has been limited,


but is expected to increase since they can be  considered relatively


fail-safe.   Deep-bed sand filters have been employed at Hanford and


Savannah River for reprocessing plant gaseous effluent filtration.


The Midwest facility has also installed a deep-bed sand filter fo.r


final filtration.


    The parameters affecting filter performance, in addition to face


velocity characteristics are;  resistance, geometrical size, pressure


drop, penetration, collection efficiency, dust and holding capacity,


particle sizes, loading, humidity, temperature, and chemical


resistance.


    Liquid Scrubbing - This method has been used in cleaning the off-


gas stream from solidification processes  (Eef. B-34 and B~35),


Scrubbers are used as initial cleaning devices and the scrubber


solutions can be recycled to the process streams.  Experience is


limited to the Idaho Chemical Processing Plant and the prototype work


conducted under the Waste Solidification Engineering Prototype


program conducted by Battelle at Hanford.


    Control Systems Evaluation - The various methods of particulars


control may be grouped in the order of increasing efficiency as


follows:


    Fair


        Group I panel filters


    Good


        a)  Group II bag filters

-------
                               B-34


        b)  Deep-bed sand or fiberglass filters


        c)  Scrubbers


    Better


        Group III bag filters


    Best


        HEPA filters


    The anticipated performance of fibrous filters can vary from 2 to


99.97% retention for 0.3 micrometer particulates and up to 100% for


10 micrometer size.  HEPA filters have consistently demonstrated


penetrations less than 0,3xlO~3   (99.97% efficiency).  However, the


overall system efficiency is highly dependent on filter integrity,


proper installation, operating conditions (particularly air flow


rates and loading), and aging characteristics.  Adequate in-place


testing is necessary since improper installation or filter damage in


shipment or installation could result in leakages up to 30%.


    A typical filter installation consists of a prefilter (roughing


filter) followed by two HEPA filters in series or by a HEPA filter


and a sand filter.  The prefilter generally has a rating of greater


than 75%  (Group III for 1.0 micrometer particles) and offers a


considerable cost savings by reducing mass loadings on the more


costly HEPA filters.  The HEPA filters are rated at a minimum


efficiency of 99.97% for 0,3 micrometer particulates.  The reported


efficiency of deep-bed sand filters for submicrometer particles is


greater than 99%.  Thus, the overall decontamination factor for the

                                              5
filter themselves should be on the order of 10 .  However, the


efficiency of the system is also  dependent upon factors such as leak

-------
                               B-35


tightness and aging characteristics.  The minimally acceptable in-

place test results would probably be in the order of 99.97% for a

typical HEPA system (it is emphasized this is a system test).  The

system would probably be assigned a credit of 99% to 99.9%, the

actual credit depending on the testing frequency.

    Capital and operating costs are highly dependent on volume flow

rates.  Based on Silverman's estimates (Ref. B-36), a HEPA filter

would have a capital cost between $200-800/1000 cfm and annual

operating costs between $100-450/1000 cfm depending upon

specifications, corrosion resistance, and other operating conditions.

A pre-filter for reducing the HEPA loading would cost between $100-

300/1000 cfm for capital cost and $15-25/1000 cfm for annual

operation.  Using approximate costs and a total flow of 60,000 cfm

the system costs would be:

                                    HEPA        Pre-filter

Capital cost (X60)               $48,000         $12,000
Annual operating costs (X60)     $18,000         $ 1,200
40-year operating costs
(present worth (§10%) .           $176,000         $11,700
Total costs (1960 prices)       $224,000         $23,700
Conversion to 1970 price -
(Miller-Stevens index 1.28)     $287,000         $30,400

    The costs of a glass-fiber or sand deep-bed filters were reported

by Silverman (Ref. B-36) to be between $2,860-5,500 capital costs per

100 cfm with associated operating cost between $400 and $800 per 100

cfm (based on only depreciation and air-moving costs).  Taking a

representative tidal flow of 60,000 cfm the cost of the deep-bed

system would be:

-------
                               B-36


    Capital cost ($3,500 x 60)              $210,000
    Annual operating cost ($800 x60)        $ 48,000
    40-year operating costs
    (present worth @10% discount)           $470,000
    Total cost (1960 prices)                $680,000
    Conversion to 1970 prices   Total       $870,000

    The reported capital cost for the sand-bed filter, discharge

stack, and associated equipment is $400,000 for the Midwest facility

(Ref.  B-33).  The corrected capital cost to 1970 for. a deep-bed sand

filter is $269,000 which agrees well with the estimate for Midwest.

    The total filter installation and 40-year operating costs at

reprocessing plants are:

                                                1970 Present Worth (§10%
    Prefilter +      filter +
    Deep-bed sand filter                             $1,187,000

    Prefilter + HEPA filter +
    HEPA filter                                      $  604,000

-------
                                    B-37

                                 REFERENCES

B-l.   Haas, W. 0., Jr., Solvent Extraction:  General Principles,
       Chapter 4, Chemical Processing of Reactor Fuels,  Flagg, ed.,
       Academic Press, New York, 1961.

B-2.   Siddall, T. H., Ill, Solvent Extraction Processes Based on
       Tri-n-butyl Phosphate, Chapter V, ChemicalProcessing of
       Reactor Fuels, Flagg, ed., Academic Press, New York, 1961.

B-3.   Benedict, M., and T. H. Pigford, Nuclear Chemical Engineering,
       McGraw-Hill, New fork, 1957.

B-4.   Stoller, S. M., and R. B. Richards, eds., Reactor Handbook
       Volume II Fuel Reprocessing., 2nd Edition, Interscience,
       New York, 1961.

B-5.   Goode, J. H., Hot Cell Evaluation of the Release of Tritium and
       85Krypton During Processing of ThO.-UQ  Fuels, ORNL-3956,
       June 1966.

B-6.   Cochran, J. A., et. al., An Investigation of Airborne Radio-
       Active Effluent from an Operating Nuclear Fuel Reprocessing
       Plant, BRH/NERHL 70-3, July 1970.

B-7.   Cochran, J. A., W. R. Griffin, Jr., and E. J. Troianello,
       Characterization of Tritium Stack Effluent from a Fuel
       Reprocessing Plant, ANS Transactions, 15, 1, Las  Vegas,
       June 1972.

B-8,   Oak Ridge National Laboratory, Siting of Fuel Reprocessing
       Plants and Waste Management Facilities, ORNL-4451, July 1970.

B-9.   Commander, R. E., et. al.» Operation of the Waste Calcining Facility
       with Highly Radioactive Aqueous Waste, ICPP, IDO-14662, June 1966.

B-10.  Magno. P. J., et. al., Liquid Waste Effluents from a Nuclear Fuel
       Reprocessing Plant, BRH/N1RHL 70-2, November 1970.

B-ll.  Oak Ridge National Laboratory, Chemical Technology Annual Progress
       Report, ORNL-4794.

B-12.  Argonne National Laboratory, Reactor Development Program Progress
       Report, ANL-7872, October 1971.

B-13.  Krumpelt, M., et. a., The Containment of Fission Product Iodine in
       the Reprocessing of LMFBR Fuels by Pyrochemical Reactions, Nuclear
       Technology, 15, September 1972.

-------
                                    B-38


B-14.  Oak Ridge National Laboratory, Aqueous Processing of LMFBR Fuels
       Technical Assessment and Experimental Program Definition,  ORNL-
       4436, June 1970

B-15.  Oak Ridge National Laboratory, National HTGR Fuel Recycle
       Development Program Plan, ORNL-4702, August 1971.

B-16.  Code of Federal Regulations, Title 10, Part 50, Appendix F,
       Policy Relating to the Siting of Fuel Reprocessing Plants  and
       Related Waste Management Facilities, November 1970,

B-17,  Russell, J. L. and F. L. Galpin, A Review of.Measured and
       Estimated Offsite Doses at Fuel Reprocessing Plants, NEA/IAEA
       Syriposium on the Management of Radioactive Wastes from Fuel
       Reprocessing, Paris, November 1972.

B-18.  Kirk, W. P., 85Krypton - A Review of the Literature and
       Analysis of Radiation Hazards, EPA, ORP, Washington, January 1972.

B-19.  Slansky, C. M., Separation Processes for Noble Gas Fission
       Products from the Off-Gas of Fuel Reprocessing Plants, Atomic
       EnergyReview, 9, 2, IAEA, Vienna.

B-20.  Merriman, H. R., M. J. Stephenson, and D. I. Dunthorn, Recent
       Developments in Controlling the Release of Noble Gases by
       Absorption in Fluorocarbons, ANS Transactions, 15s 1, Las  Vegas,
       1972.

B-21.  Keilholtz, G, W., Krypton-Xenon Removal Systems, Nuclear Safety
       12, 6, November-December 1971.

B-22.  Nichols, J. P. and F. T. Binford, Status of Noble Gas Removal
       and Disposal ORNL-TM-3515, August 1971 ,

B-23.  Bendixsen, C, K., and G. F. Offutt, Rare Gas Recovery Facility
       at the Idaho Chemical Processing Plant, Idaho Nuclear Corp.,
       IN-1221, April 1969.

B-24.  Benixsen C, L., and F. 0. German, Operation of the ICPP Rare
       Gas Recovery Facility During Fiscal Year 1970, Allied Chemical
       Corp., ICP-1001, October 1971.

B-25.  Stephenson, M. J. J. R. Merriman, and D. I. Dunthorn, Experimental
       Investigation of the Removal of Krypton and Xenon from Contaminated
       Gas Streams by Selective Absorption in Fluorocarbon Solvents:
       Phase I Completion Report, Union Carbide, K-1780, August 1970.

-------
                                    B-39


B-26.  General Electric Company, Midwest Fuel Recovery Plant,
       Applicant's Environmental Report, Supplement 1, NED 14504-2
       November 1971 -

B-27.  Environmental Protection Agency, The Separation and Control of
       Tritium State-of~the-Art Study, Pacific Northwest Laboratories.
       BMI, April 1972.

B-28.  ORNL-4794 Chem. Tech. Div,, Annual Progress Report Period
       Ending March 31, 1972.

B-29.  Personal Communication, 0. 0. Yarboro, Oak Ridge National
       Laboratory Chemical Technology Division ,

B-30.  Pence, D. T., P. A. Duce, and W. J. Maeck, Application of Metal
       Zeolites to Nuclear Fuel Reprocessing Plant Off-Gas Treatment,
       ANS Transactions, 15, 1, Las Vegas, 1972.

B-31.  Wilhelm, J. G., and H. Schuttelkopf, An Inorganic Adsorber
       Material for Off-Gas Cleaning in Fuel Reprocessing Plants,
       12th AEG Air Cleaning Conference, Proceedings to be published
       August 1972.

B-32.  Burchsted, C. A., and A. B. Fuller, Design Construction, and Testing
       of High Efficiency Air Filtration Systems  for Nuclear Application,
       ORNL-NSIC-65, January 1970.

B-33.  Juvinall, R. A., R. W. Kessil, and M. J. Steindler, Sand-Bed
       Filtration of Aerosols;  A Review of Published Information on
       Their Use in Industrial and Stomic Energy  Facilities. Argonne
       National Laboratory, ANL-7683, June 1970.

B-34.  Lakey, L. T. and B. R. Wheeler, Solidification of High-Level
       Radioactive Wastes at the Idaho Chemical Processing Plant,
       to be presented at NEA/IAEA Symposium on the Management of
       Radioactive Wastes from Fuel Reprocessing, November 1972.

B-35.  McElroy, J. L., A. G. Blaservits, and 1. J. Schneider, Status
       of the Waste Solidification Demonstration  Program, Nuclear
       Technology, 12, September 1971.

B-36.  Silvernan, L., Economic Aspects of Air and Gas Cleaning for
       Nuclear Energy Processes, Proceeding of a  Conference on the
       Disposal of Radioactive Wastes. Vol. I, IAEA, Vienna, 1960.

B-37.  U.S. Atomic Energy Commission, Environmental Statement
       Plutonium Recovery Facility Rocky Flats Plant Colorado,
       WASH-1507, January 1972.

-------
                                   B-40

B-38.  Report of the Committee to Study the Establishment of Plants
       or Facilities for the Recovery of Nuclear Fuel and the
       Storage of Waste Nuclear Materials, 1972.

B-39,  Hearings on Midwest Plant before Illinois Pollution Control
       Board.

B-40,  Martin, J. A., Jr., Calculations of Doses in 1971 Due
       to Radionuclides Emitted by Nuclear Fuel Services Fuel
       Reprocessing Plant, Radiat Data Rep. 14:59-76 (February 1973)-

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           APPENDIX C










      RADIOLOGICAL DOSE AND




HEALTH IMPACT CONVERSION FACTORS

-------

-------
                                      C-l

I.    INTRODUCTION

          Radioactive materials released into the environment from nuclear

      fuel reprocessing become dispersed in the surrounding media (air,  water,

      etc.) and ultimately may produce health effects in man.  The impact

      of a given radionuclide release on the population surrounding a source

      is assessed here in terms of three factors: (1) a dilution factor  to

      calculate the concentration of the released activity in the medium of

      interest, (2) a medium concentration to dose conversion factor, and

      (3) a risk factor which relates the likelihood of a given biological

      effect to an absorbed dose of one  rad. 1  These factors are discussed

      below for tritium, krypton-85, the radioiodines, and alpha-emitting

      transuranics (such as plutonium).

II.   PATHWAYS

          Releases of radionuclides from a nuclear fuel reprocessing plant

      can occur by venting through an exhaust stack to the atmosphere or

      drainage to a nearby waterway.  The principal pathway of concern in

      assessing the health impact due to nuclear fuel reprocessing plant

      operations is the atmospheric pathway because such releases can

      become dispersed in any direction and lead directly to radiation

      exposure to man.  The significance of the water pathway is expected to

      be quite small for future reprocessing plants because presently proposed

      designs do not plan on any releases to waterways.

          Atmospheric dispersion of radioactivity has been discussed by  a

      number of authors (Ref. C-l, C-2)  and for the case of interest here, a
       Rad - The unit of energy imparted to matter by ionizing radiation
      and equal to  ,01 J/kg in any medium.

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                                C-2





gaussian plume diffusion model is usually assumed to be the best choice.




At distances relatively close to a source (2-3 km) this model can




predict the air concentration of krypton-85, for example, within a




factor of two or three.  Its applicability at longer distances depends




upon the local weather conditions at the time of radioactive release




and the topography.  For unstable atmospheric conditions, it may be




reasonably accurate as far as 10 km from a source.  Since the average




radionuclide concentrations in air around sites have been calculated




for distances as far as 80 km (SO miles) from the source, it is obvious




that the validity of the atmospheric transport model used is an important




limitation.  However, the point here is to examine the general case




and provide an overall index of health risk; the risk from a particular




plant will depend on the details of the local meteorological situation.




For worldwide distribution of gases, uniform dispersion was assumed in




determining air concentrations.




    An important radionuclide pathway for man is the direct contam-




ination of foodstuffs - particularly milk.  For iodine-131, this




pathway has been studied extensively by several authors including




Garner (Ref. C-3) and Bryant (Eef. C-5, C-5).  Long-term buildup of




the isotope iodine-129 may be important due to its half-life of




17xl06 years but appreciable buildup has not been documented




in the literature.  The short-term behavior of iodine-129 has been




considered, however (Ref. C-6), and health risk estimates are given




for both iodine-129 and iodine-131.

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                                       C-3




III.   MEDIA CONCENTRATION TO DOSE CONVERSION FACTORS




            Organ or total body dose estimates are critically sensitive to




       assumptions concerning the route of uptake, the amount of radioactivity




       inhaled or ingested daily, the fraction of activity retained in the- organ




       of interest, and the residence -time of the activity in various parts of




       the body.  The final necessary elements entering into the dose computations




       are the physical considerations of organ mass and radionuclide distri-




       bution within the organ.  In the present state of the art, the complex-




       ities of the radionuclide distribution within organs are nearly always




       circumvented by assutalng a uniform depositon.  Information concerning




       the other inputs is based mainly on empirical evidence, largely gathered




       from fallout studies and medical investigations.  In order to reduce




       the number of variables to be considered in dose calculations, the




       International Commission on Radiological Protection (ICRP) has postu-




       lated a "standard man;" i.e., a model.system having standardized biological




       parameters based on either average values or best estimates as listed




       in the scientific literature.  The standard man is a hypothetical adult




       industrial worker and it is not clear to what extent parameters so de-




       fined are applicable to an environmentally exposed population.




            For particular radionuclides, the sensitivity of certain age groups




       may be the limiting factor.  In the case of iodine-131, the Federal




       Radiation Council (Ref. C-7) has defined children as the most sensitive




       population group and, therefore, the biological parameters used in




       these media to dose conversion factors are not based on standard man.




       Rather, models appropriate for children's thyroid glands and thyroid




       metabolism have been used.  For the other radionuclides considered here,

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                               C-4





little is known concerning differences between adults and children.




Such differences are seldom considered in the literature.  This being




the case, the conversion factors listed in the subsequent sections, while




adequate, must be considered only as first order approximations and




not as definitive dose estimates from environmentally distributed




radionuclides.



     Media concentration-to-dose conversion factors are defined below




for krypton-85» tritium, iodine 131, iodine~129, and some of the    s




actinides.  Other radionuclides are not considered likely to cause




significant environmental exposures of the population based on the




control technologies discussed in appendix B.




A.   Rangeof Expected Doses from Krypton-85 Exposure




     Since krypton-85 is not metabolized, exposure to humans is limited




to external beta and gamma rays and, to a much lesser extent, the dose




due to krypton-85 dissolved in body fluids.  The health risk from




krypton-85 is further limited by the fact that 99% of the decay energy




is dissipated by beta rays which have no potential for deep penetration.




     Four target organs are considered for these dose and risk estimates:




total body, gonad, lung, and skin.  In each case it  can be shown that




only one type of exposure need be considered, the other contributing




an insignificant fraction of the dose.




     Kirk (Eef. C-8) has recently reviewed the literature on krypton-85




dose and established relationships between the concentration in air of




krypton-85 and various organ doses.  A review of these results show




whidh radiations and source locations are important.  For the whole




body, dose and risk estimates can be. based on a consideration of

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                               0-5




external photon exposures, i.e.,  gamma rays and bremsstrahlung.  For




genetic risk calculations the gonadal dose, in the case of males, is




from exposure from external photons; while for females, the whole body




dose estimate can be used.  Dose estimates for the lung are based on




internal beta dose plus the total body gamma-ray dose.  Skin dose and




risk estimates are based on the dose delivered by external beta




radiation after making an appropriate allowance (0,25) for the shielding




provided by clothing and the nonviable epithelium.




B.   Range of Expected Doses from Tritium Exposure




     Dose estimates from tritium.exposure have been based on the




assumption that the isotope is contained in body water (Ref. C-9).




However, chronic exposure to environmental tritium has been shown to




lead to incorporation into organic molecules from which tritium is lost




at a slower rate than from body water (Ref. C-10, C-ll).  If it is




Assumed that, under equilibrium conditions, all body hydrogen  (7.0 kg in




standard man) is uniformly labelled, a sustained concentration of 1 yCi/1




body water would lead to a body burden of 63 uCi, as opposed to 43 yCi if,




as in the ICRP model only, distribution in body water alone is considered.




Ivans (Ref. C-10) found that tritium was not, in fact, uniformly




distributed through deer tissues and, assuming that his observed factors




are applicable to man$ has calculated that the body burden carried by




standard man at a sustained concentration in body water of 1 yCi/1 would




be 60 yCi, i.e., higher by a factor of 1.4 than that based on  the ICRP




model.  While Evans* factor has been adopted in some dose calculations,




a factor of 1.5 (63/43), although only marginally different, may be a




more appropriate value to use for calculations in man and, therefore, it




is used here.

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                               C-6

     Since it is apparent that, under chronic exposure conditions,

tritium may become incorporated into the genetic material (DMA), it has

been suggested that the relative biological effectiveness of tritium

in terms of genetic effects may be greater than unity as a result of DNA

degradation from transmutation and recoil processes in addition to that due

to absorbed energy from ionization processes due to beta emissions.  How-

ever, from both experimental (Ref. C-12) and theoretical (lef. C-13)

considerations, it has been concluded that it is the absorbed dose to

mammalian cell nuclei from incorporated internuclear tritium which deter-

mines quantitatively the degree of effect (Ref. C-24).  The assumption

made in these calculations that the appropriate value for quality factor

for tritium dose equivalent estimation is 1.0 as recently adopted by the

National Council on Radiation Protection and Measurements (NCRP)  (Ref. C-14),

     A sustained concentration of  1 yCi tritium per liter of -body water

would thus be equivalent to a specific activity (assuming uniform

labelling of all body hydrogen) of 9x10  3 yCi tritium/g hydrogen, and

would deliver an annual dose to body tissues of approximately 100 mrem.

     The concentration of tritium in body water resulting from  exposure

to tritium in air is obtained by diluting the daily intake of tritium

by inhalation into the 43 liters of body water with a biological half-

life of 12 days.  This amount of tritium is doubled to account  for

absorption of tritium through the skin.  This leads to an annual 'dose

of 1,7x106 rem2 for an air concentration of 1  Ci  tritium/cm^.
    the case of beta and gamma rays emitted by fuel reprocessing plant
effluents, the quality factor is one and the dose  equivalent  in rents
is identical to the dose in rads.  Where the effects of such  effluents
are considered in this report, doses are expressed in rem units and
biological effects are presented on a per rem basis.
KM - The rem represents that quantity of radiation that is equivalent—
in biological damage of a specified sort - to 1 rad of 250 kVp x rays.

-------
                                C-7





C.   Range of Expected Doses from Iodine-129 and I6difle-131 Exposures




     Atmospheric releases of iodine from fuel reprocessing may result




in an accumulation of iodine-129 and iodine-131 in the thyroid glands




of persons living in the surrounding area.  The pathway potentially most




hazardous to man for isotopes of iodine is the grass-cow-milk chain,




particularly in cases where the milk is not diluted with uncontaminated




supplies.  Direct deposition on foliage will be the only significant




route of contamination of edible herbage for iodlne-131 and is likely to




be the most important for iodine-129, at least over the duration of




plant operation.




     Because of the long half-life of iodine-129, recycling through




the soil should be considered.  In organically rich soils, the iodine




will be strongly bounded to the soil, but it will be leached rather rapidly




from other types of soils.  In any case, plants will incorporate




iodine-129 in ratio proportional to the amount of natural iodine-127




available.  The actual amount of iodine-129 incorporated will depend on




the location of the reprocessing plant, and the specific activity of




the iodine-129  (curies of iodine-129 per gram of iodine) in each component




of the terrestrial pathway will change as a function of time as build-




up in the soil increases.  At any given time the specific activity in




the ecological chain will be somewhat less that the specific activity of




the iodine-129 in the air.  In many cases the specific activity will be




much less because of the large stable iodine reservoir in soils and other




parts of the terrestrial pathway.

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                               C-8





     When considering the exposure of individuals to iodine isotopes via




the grass-cow-milk chain, the population potentially at greatest risk




is young children consuming fresh milk (Ref, C-15, C-16).  From the data




of Durbin, et al. (Ref. C-1J) the average daily intake of whole cows milk




by U.S. children over the first year of life is about 760 ml.  Appropriate




representative data to define the relationship between the amount of




iodine ingested by a 6-month-old child and its concentration in the




critical organ, the thyroid gland, are (Ref. C-6); thyroid weight, 1.8 g;




fraction of ingested iodine reaching critical organ, 0.35; and biolog-




ical half-life of iodine in thyroid, 23 d.  Equivalent data for adults,




appropriate to the calculation of average population doses, are: daily




milk consumption, 500 ml (Ref. C-17); thyroid weight, 20 g; fraction




ingested reaching critical organ, 0.3; and biological half-life in




thyroid, 138 d (Ref. C-6).  Use of these values yields an annual dose




to the adult thyroid of 0.29 mrem for iodine-131 or 1.9 mrem for iodine-




129 for a concentration of 1 pCi/1 of each isotope.  The corresponding




annual doses to the thyroid of children whose daily consumption of milk




during the first year of life contains 1 Ci/1 of the respective




radionuclldes are 4.3 mrem for iodine-131 and 6.3 mrem for iodine-129.




The media conversion factors presented in table C.I are also derived




from considerations discussed in references C-6 and C-7, an assumed




grazing area for a dairy cow of 80 m2 per day and an iodine deposition




velocity of 0.5 cm/s (Ref. C-5).

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                                      C-9
                               TABLE C.I

            Milk Concentrations of Iodine-131 and Iodine-129 from
             Given Input Concentration and Corresponding Doses
      Input Concentration
     of Respective Nuclide
                                            Milk Concentrations
 Iodine-131
Iodine-129
       o
1 pCi/m   (ground surface)

1 pCi/m3  (air)
1 pCi/m   (air)
1 pCi/m   (air)
 0.20 pCi/1
 6.2xl02 pCi/1
0.28 pCi/1
2.4x103 pCi/1
 Annual Dose to Child Thyroid

   Iodine-131          Iodine-129


 2.7 rem             15 rera

Annual Dose to Adult Thyroid	
                                        Iodine-131
                    Iodine-129
  .18 rem
 4.6 rem
            Estimates of the specific activity (yCi iodine-129/g total iodine) in

       the thyroid gland corresponding to an annual dose of 1 rem are, for an

       adult, 2.3, and for a 6-month child, 4.1 (lef, C-6).  Adoption of a

       value of 0.44 rem/yr as the dose delivered to a thyroid gland containing

       1 yCi iodine-129/m3 total iodine would thus appear to be a reasonable

       estimate for all cases.  The mean concentration of stable iodine in the

       atmosphere is given as 0.2 Ug/m3 (Ref. C-18).  Using this value, it can

                                                               0
       be shown that an air concentration of 1 pCi iodine-129/m  would lead to

       an annual thyroid dose of 1.8 rem.  In areas where the atmospheric

-------
                                C-10




concentration of stable iodine-127 is low, the dose could be up to




40 times higher; this upper limit on the dose is set by the amount




of stable iodine-127 released in fuel reprocessing.  Thus, for




adults the higher value presented in table C.I of 4.6 rem/yr




per pCi/m3 of iodine-129 in air is selected for use in this study.




D.   Range of Expected Doses from Plutonium-239 and Other Actinide Exposures




     The potential risks from Inhalation of plutonium-239 depend on whether




the plutonium is in a soluble or an insoluble form.  Present experience in-




dicates that, in the case of fuel reprocessing, the plutonium will be pre-




sent in the environment in a relatively insoluble form and the present




dose estimates are based on this assumption.  There is also evidence that a




considerable fraction of plutonium-239 inhaled in insoluble form is trans-




located, largely to the bronchial and mediastinal lymph nodes (Ref. C-19).




Since the risk to be pulmonary region depends upon both the amount of




plutonium in the organ and its microdistribution, the region containing




the largest amount of plutonium may not be the region at greatest risk.




Particularly, since the relative sensitivity of the various cell types




encountered has not been established, the dose to the lung from inhaled




particules is calculated on the basis' of an average dose to the entire




pulmonary region for this report.  In the case of alpha emitters, such




averaging is obviously inappropriate if there are only a few particles




present.  ICRP Publication No. 6 (Ref, C-20) recognizes this and states, "in




the case of the lung, an estimate of the dose equivalent to the critical




tissue determined merely by the product of quality factor and mean dose




may be greatly in error, but further experimental evidence is needed before




a better estimate can be made."

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                                e-ii

     In this report, dose estimates are based on the new ICRP lung model

(lef. C-21, C-22),  The biological half-life of material in the lung

(pulmonary region) is assumed to be 1,000 days.  Using this model,

sustained exposure to an air concentration of 1 pCi/m^ of

plutonium-239 in insoluble form would lead to a dose rate of 12 rem per

year in the pulmonary region.  It is assumed that the risk to this

region is representative of the total risk to the lung.

     Media-dose conversion factors for other actinide radionuclides

are related to the plutonium-239 conversion factor by taking into

account the effective energy absorbed per disintegration of each

radionuclide and the physical half-life as given in IC1P Publication

Nos. 2 and 6 (Ref C-9, C-20).  Table C.2 gives the conversion factors

used in this study for several actinides relative to plutonium-239.



                          TABLE C.2

              Actinide Air-Dose Conversion Factors
                 Relative to Plutonium-239

Radionuclide
Pu-238
Pu-239
Pu-240
Pu-241
Am-241
Cm-242
Qn-244
a
Relative Conversion Factor
1
1
1
0.001
0.25
0.17
0.33
 PI"tonium-239 Conversion Factor - (12xl06 rem/yr)/(l PCi/m3)

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                              C-12





E.  Summary




     Table C.3 summarizes the media-dose conversion factors presented




in this section.  Conversion factors are expressed in terms of




rem/yr resulting from continuous exposure to concentrations




expressed in yCi/cm3 of air.






                    TABLE  C.3




     Summary of Media-Dose Conversion Factors
Radionuclide Critical Organ
Kr-85 Whole body
Gonads (female)
Gonads (male)
Lung
Skin
H-3 Whole body
1-129 Infant thyroid
Adult thyroid
1-131 Infant thyroid
Adult thyroid
Pu-239 Lung a
a See paragraph IV D regarding
I?. DOSE-RISK CONVERSION FACTORS
Conversion Factor
(rem/yr) /(yCi/cm3 air)
1.5x10^
1.5x10^
2.0x10^
3.0x10**
50.0x10^
1.7xl06 ot
100 ( rem/yr) /(pCi/em3 water)
15. 0x10 1 2
4.6xl012
2.7xl012
0. 18x10 l 2
12. 0x10 12
consequences of soluble form of plut

     Assumptions made in predicting radiation-induced health effects



from nuclear fuel reprocessing are given in this section.  Consistent



with recommendations made in the recent (November 1972) National

-------
                                C-13




Academy of Sciences Committee on Biological Effects of Ionizing




Radiation (BEIR) report (Ref. C-23), the health risks presented in this




report are based on an assumed linear relationship between absorbed




dose and biological effects and that any increased risk is in




addition to that produced by natural radiation; i.e., no threshold




exists.  It is further assumed that health effects that have been




observed at dose rates much greater than those likely to be encountered




around fuel reprocessing plants are indicative of radiation effects at




lower dose rates.  Only insofar as any biological repair of radiation




damage from low dose rate radiation is neglected do the BEIR




health risk estimates represent upper limits of risk.  In most cases




the risk estimates are based on relatively large doses where cell




killing may have influenced the probability of delayed effects being




observed.  The BEIR risk estimates used in this report are neither




upper nor lower estimates of risk, but simply the "best available."




     As the BEIR report points out, a nonthreshold linear relationship




hypothesis is not in itself sufficient for the prediction of health




risk.  It is also necessary to assume that all members of the exposed




population have equal sensitivity to the radiation insult so that the




expression of health risk is independent of how individual exposures




are distributed.  This requirement is not wholly satisfied.  As




documented in the BEIR report, differences in sensitivity do exist;




for example, children are more radiosensitive than adults.  There are




two considerations, however, which help validate the application of




available mortality data to a consideration of health effects from

-------
                                C-14


fuel reprocessing.  Some of these data (those taken from Hiroshima and


Nagasaki) reflect, to a limited extent, exposure of a relatively


heterogeneous population.  More Importantly, even though the number of


health effects will be dependendent on the exact makeup of the


populations at risk, the relative order of importance of the various


pathways of exposure will not be very sensitive to the population


characteristics near a given fuel reprocessing plant.  Finally, it


should be pointed out that the health risk estimates made here assume


that the expected radiation effects are independent of other environ-


mental stresses, which may be either unique to the population surrounding


fuel reprocessing facilities or unique to the exposed groups considered


in the BEIR report.


     The numerical risk estimates used in this appendix are primarily


from the BEIR report.  What must be emphasized is that though these


numbers may be used as the best available for the purpose of risk-cost


benefit analyses, they cannot be used to accurately predict the number


of casualities,  for a given dose equivalent, the BEIR report estimates
                     i

a range for the health impact per million exposed persons.  For example,


the BBIR results from a study of the major sources of cancer mortality

                           »3                                          g
data yield an absolute risk  estimate of 54-123 deaths annually per 10


persons per rem for a 27-year followup period.  Depending on the details


of the risk model used, the BEIR Committee's relative risk  estimate is
 ~t
 Absolute risk estimates are based on the reported number of cancer deaths
 per rad that have been observed in exposed population groups, e.g.,

 Hiroshima, Nagasaki, etc.


 Relative risk estimates are baaed on the percentage increase of the
 ambient cancer mortality per rem.

-------
                              C-15






160-450 deaths per 106 persons per rem.  It is seen that the precision




of these estimates is at best about a factor of 3-4, even when applied




to sample populations studied on the basis of the same dose rates.  The




application of the BEIR risk estimates to exposures at lower dose rates




and to population groups more heterogenous than those studied increases




the uncertainty in the risk estimates.  Considering the limitations




of presently available data and the lack of an accepted theory of




radiocarcinogenesis, emphasis should be placed on the difference in




risk estimates between the various procedures and countermeasures




discussed in this report rather than on the absolute numbers.  Where the




absolute numbers must be used for risk-cost-benefit balancing, it




should be remembered that these risk estimates are likely to be re-




vised as new information becomes available.  Notwithstanding these




disclaimers, it is also pertinent to note that we are in a better




position to evaluate the true risks and the accompanying uncertainties




from low levels of radiation than from low concentrations of other




environmental pollutants which might affect populations in the vicinity




of a fuel reprocessing plant.




A.   Dose-Risk Conversion Factors for Krypton-85




     1.  Total Body Dose-Risk




         The BEIR report calculates the cancer mortality risk




(including leukemia mortality) from whole body radiation by two quite




different models.  The absolute risk model predicts about 100 cancer




deaths per 10  person-rem while the relative risk model predicts between




160 and 450.  An average cancer mortality of 300 annually per 106




person-rem would seem to be an appropriate mean for the relative risk

-------
                              C-16



model.  The average of the absolute and relative risk models is 200,




which is close to the estimates of cancer mortality risk listed as




"most likely" by the Committee,  Cancer mortality is not a measure of




the total cancer risk, which the Committee states is about twice that




of the yearly mortality.




     The computation of health risk from continuous krypton-85




total body exposure used in this report is the one appropriate for




total body irradiation.




        Estimated Cancer Risk from TotalBody Irradiation  .




     Cancer mortality = 200 deaths per year for 106 person-rem




     annual exposure.  Total cancers = 400 cancers per year for




     10  person-rem annual exposure.




     2.  Gonadal Dose-Risk




         The range of the risk estimates for genetic effects set forth




in the BEIR report is so large that such risks are better considered on a




relative basis for different exposure situations than in terms of




absolute numbers.  The range of uncertainty for the "doubling dose"




(the dose required to double the natural mutation rate) is 10-fold




(from 20 to 200 rad); and because of the additional uncertainties in




(1) the fraction of presently observed genetic effects due to back-




ground radiation, and (2) the fraction of deleterious mutations




eliminated per generation, the overall uncertainty is about a factor of




25.  The total number of individuals showing very serious genetic




effects such as congenital anomalies, constitutional and degenerative

-------
                              C-17




diseases, etc., is estimated at somewhere between 1,800 and 45,000 per




generation per rad of continuous exposure; i.e., 60-1,500 per year if




a 30-year generation time is assumed.  This level of effect will not be




reached until after several generations of exposure; the risk to the




first generation postexposure is about a factor of 5 less.




     The authors of the BEIR report reject the notion of "genetic death"




as a measure of radiation risk.  Their risk analysis is in terms of




early and delayed effects observed post parturn and not early abortion,




stillbirths or reduced fecundity. . Because of the seriousness of the




genetic effects considered here, e.g., mongolism, the emotional and




financial stress would be somewhat similar to death impact.  Indeed,




many of the effects described are those which lead directly to infant




mortality (fetal mortality is excluded).  For the purposes of this




report this class of genetic effects will be considered on the same




basis as mortality data,




            Estimated Serious Genetic Risk




          from ContinuousGonadal Irradiation




     Total risk = 300 effects per year for 106 person-rem annual




     exposure.




         Less serious genetic effects have also been considered by the




BEIR Committee.  These have been quantified under the categorgy "unspeci-




fied ill health."  The Committee states that a continuous exposure of




one rem per year would lead to an increase in ill health of between




3 and 30 percent.

-------
                                C-18

     3.  Lung Dose-Risk

         Due to the insufficient data for the younger age groups,

estimates of lung cancer mortality in the Bill report are only for

that fraction of the population of age 10 or more.  For the risk esti-  '

mate made below, it is assumed that the fractional abundance for lung

tumors is the same for those irradiated at less than 10 years of age

as it is for those over 105.  On an absolute risk basis lung cancer mortality

is about 26 deaths per annum per 10 ^ persons irradiated continuously

at a dose rate of 1 rem per year.  This is a minimum value.  The BEIR

report states that .the absolute risk estimates may be too low because

observation times for exposed persons are still relatively short compared

to the long latency period for lung cancer.  Furthermore, lung cancer

risks calculated on the basis of relative risk would be larger.  For

the risk estimate made here, it is assumed that the ratio of absolute risk

to the average relative risk is at least a factor of 0.5, i.e., the

same ratio as in the case of total body irradiation discussed above.

                     Estimated Lung Cancer Risk                          '

                    fromContinuous Lung Irradiation

     Lung cancer mortality = 50 deaths per year for 10  person-rem

     annual exposure.

     4.  Skin Dose-Bisk

         There is no doubt that the dose to the skin delivered by

krypton-85 exceeds by about two orders of magnitude the insult to other

organs.  However, epidemiological evidence of any real risk from such
 5An absolute risk estimate is not very sensitive to the inclusions of this
 assumption since lung cancer incidence is very small in the young.

-------
           •      ;             C-19



insults at the dose levels ebnsidered here is nonexistent.  This is



not to say that the linear dose-effect, assumption does not,hold for


skin cancer but rather that the.BEIR Committee found that from the


extensive evidence, they examined "numerical.estimates of risk at low



dose levels would not seem to be warranted."  The authors of this report



concurred with the BEIR report.  However, rather than defining a


zero risk per rad for any radiation insult from krypton-85, an upper


limit of risk is proposed.  It should be noted that skin cancers are


rarely fatal and usually not very debilitating.  The estimated


upper risk for continuous exposure is:

                                               C                C
     Skin cancer - upper limit = 3 skin cancers" per year for 10° perspn-



     rem annual exposure.


B.   Dose-Risk Conversion Factors for Tritium


     1.  Total Body Dose-Risk                                         :


         The somatic dose-effects from tritium are not expected to be


unique.  Risk estimates for total body irradiation are based on the



same information reviewed in Sectioii. A for krypton-85 total body


exposure.


     2.  Gonadal Dose-Risk


         The genetic risk from tritium per unit gonadal dose is expected


to be the same as for the beta and gamma radiation from other isotopes.


Some experiments with bacteria (Ref. C-24) have shown that the  location



of a tritium atom on a particular DNA base can enhance the mutation rate,
 5Assuming  30 years at risk exposure  and  that  the  incidence of  skin  cancer

 will be  10% of  all radiation-produced  cancers  except  leukemia, breast,

 lung,  G.I. tract, and bone cancers.

-------
                                 C-20
However, if it is assumed that tritium labeling is a random phenomenon,



the percentage for such locations that are specifically labeled will be



extremely small at the exposure levels considered here.  Therefore, the



risk estimates for gonadal .irradiation from krypton-85 listed in section



'A are also appropriate for estimating the genetic risk from tritium .



exposures,



C.  Dose-Risk Conversion Factors for lQdine-129 andIodine-131



    Iodine is concentrated in the human thyroid.  Therefore, the insult



from radioiodines is important only for the thyroid.  The dose to other



organs is orders of magnitude less.  Two health effects follow high



level exposures of thyroid tissue to ionizing radiation: benign neoplasms



and thyroid cancer.  Though the former are a more common radiation



effect, only the risk from cancer is considered here.



    While children are particularly sensitive to radiation damage to



their thyroid glands, thyroid cancer is not usually a deadly disease



for persons in younger age groups but mortality approaches 25% in



persons well past middle age.  It is not presently known if the radi-



ation-induced cancers which are more frequent for persons irradiated



early in life will follow the same patterns of late mortality.



    The BEIR report provides risk estimates only for morbidity (not



mortality) and only for persons under 9 years of agej i.e., 1.6-



9.3 cancers per 10^ person-rem years.  From the Hiroshima data and



other studies it would appear that for persons over 20 years old the



radiation-induced thyroid cancer incidence is much lower and may



approach zero.

-------
                                C-21

     Since information in the BEIR report Is not sufficient in itself

to estimate the cancer incidence from continuous exposure, tentative

risk estimates for this study are also based on risk estimates described

in ICRP'Publication No. '8 (Ref. C-25) as well as the mean of the BEIR

Committee's various estimates of incidence per rem.  Infants and fetuses,

composing,approximately 2.5 percent of the population, are, of course, the

most sensitive group.  For this group about 150 thyroid cancers may accrue

annually per 10  persdn-rem annual exposure.  For the approximately

401 of the populations that is in the 1-19 year age group it is assumed

that the incidence is a factor of about 4 less, and that for the balance

of the population, it is a factor of 30 less,

      EstimatedThyroid Cancer Risk from Radioiodines

     Morbidity for less than age 1:
     150 cases per year for 10  person-rem annual exposure.

     Morbidity for less than age 20:
     35 cases per year for 10  person-rem annual exposure.

     Morbidity for more than age 20:
   ,  5 cases per year for 106 person-rem annual exposure.

     It is unlikely that the annual mortality from this cancer would be

much larger than 25% of the rate of incidence.  As for other radiation

effects, a true measure of the risk from thyroid cancers would-be life

shortening, but insufficient mortality data prevents such an approach.

D.   Dose-Risk Conversion Factors for Plutonium and other:Actinides

     The lung cancer mortality risk discussed in Section A is the best

available information on the consequences of lung exposure.  While it

is based on mortality information from miners exposed to alpha emissions

from particulates as well as more conventional dose-risk data, it is

-------
                              C-22 .






probably not really adequate for describing the risk from inhaled




plutonium.  There is good evidence that a fraction of such particulates




are cleared from the lungs and relocated into the respiratory lymph




nodes.  The organ dose received by a lymph node in this case is not




really known, but is probably on the order of 50 times the average




dose to lung tissue.  The ICKP does not consider these highly irradiated




nodes to be the organ at maximum risk and preliminary results of animal




experiments would tend to confirm their judgment.  However, it is not




a settled question.




     Even if the lung is the critical organ for such exposures, there




is little reason to believe that the average lung dose, presently used




in health-risk analyses, is really relevant to estimating the risk from




air-borne particles.  Estimates of the actual dose from discrete sources




of alpha radiation are subject to large variability simply because




little is known about the volume over which the energy deposition takes




place.  Even though as many as 4x10  particles (0.2 Pm diameter) of




plutonium-239 are required to deliver an average energy deposition of




1 rad to the lung, the dose is not evenly distributed; only about 0.2%




of the lung volume absorbs the emitted energy.  Health risk estimates




based on dosimetry are probably unwarranted under these circumstances




and use of a body burden approach to health-risk assessments would




appear to be a more likely route to Success,  Unfortunately, experiments




allowing this approach are not yet complete.  Therefore, the lung




cancer risk estimate for exposure to actinides for purposes of this




study is as given in Section A for uniform lung exposure.

-------
                              C-23





     The dose conversion factor for a soluble form of plutonium will




differ from that presented for insoluble plutonium for a given air




concentration.  However, the associated risk (expressed in effects




induced per unit air concentration) resulting from the soluble form




is expected to be the same order as for the insoluble form as




analyzed here.

-------
                                  C-24


                                REFERENCES

 C-l    U.S.  Atomic Energy Commission, Meteorologyand Atomic Energy - 1968;
        Slade, D.H., Editor, Division of Reactor Development and Technology,
        July, 1968.

 G-2    Turner, D.  Bruce (ESSA), Workbook, of Atmospheric Dispersion Estimates,
        U.S.  Department of Health," Education, and Welfare, Public Health
        Service, Consumer Protection and Environmental Health Service,
        National Air Pollution Control Administration, Cincinatti,
        Revised 1969.

 C-3    Garner, R.J., and Russell, R.S., Isotopes of Iodine,
        Radioactivity and HumanDiet, ed.  R.S. Russell, Pergamon Press, 1966.

 C-4    Bryant, P.M., Data for Assessments Concerning Controlled and Accidental
        Releases of i31I and 137Cs to Atmosphere,  Health Physics, Vol. 17,
        pp.  51-57,  1969..

 C-5    Bryant, P.M., Derivation of Working Limits for Continuous Release
        Rate of Iodine-131 to Atmosphere in Milk Producing Area,
        Health Physics, Vol. 10, pp. 249-257, 1964.

 C-6    Bryant, P.M., Derivation of Working Limits for Continuous Release
        Rates of 129I to Atmosphere, HealthPhysics, Vol. 19
        pp.  61.1-616, 1970.

 C-7    Federal Radiation Council, Background Material for the Development
        of Radiation Protection Standards,  Staff Report No. 1, May 13, 1960.

 C-8    Kirk, W. P., Krypton-85 - A Review of the Literature and an Analysis
        of Radiation Hazards,  Environmental Protection Agency, Office of
        Research and Monitoring, Eastern Environmental Radiation Laboratory,
        January, 1972.

 C-9    International Commission on Radiological Protection (ICRP), Recommenda-
        tions of the International Commission on Radiological Protection, Report
        of Committee II on Permissible Dose for Internal Radiation,  ICRP Publica-
        tion 2, Pergamon Press,  1959.

C-10    Evans, A.G., New Dose Estimates from Chronic Tritium Exposures,
        Health Physics. Vol. 16, pp. 57-63, 1969.

C-ll    Koranda, J.J., and Martin,  .R., Persistence of Radionuclides at Sites
        of Nuclear Detonations,   BiologicalImplications of the Nuclear Age,
        U.S. Atomic Energy Commission Symposium Series No. 5, 1965.

-------
                                      C-25

C-12   Bond, V.P., and Feinendegen, L.E., Intranuclear % Thymidine, Dosimetrie,
       Radiobiological and Radiation Protection Aspects,  Health Physics, Vol. 12
       pp. 1007-1020, 1966.

C-13   Tompkins, P.C., Environmental Radiation Protection Criteria and
       Tritium,  Paper presented at Tritium Symposium, Las Vegas, Nevada,
       August, 1971.

C-14   National Council on Radiation Protection and Measurements, Basic
       Radiation Protection Criteria,"CEP Report No. 39, January, 1971.,

C-15   Federal Radiation Council, Background Material for the Development
       of Radiation Protection Standards,  Staff Report No. 2, May 13, 1960.

C-16   Federal Radiation Council, Background Material for the Development
       of Radiation Protection Standards,  Staff Report No. 5, July, 1964.

C-17   Durbin, P.W., Lynch, J., and Murray, S., Average Milk and Mineral
       Intakes (Calcium, Phosphates, Sodium and Potassium) of Infants in the
       United States from 1954-1968:  Implications for Estimating Annual
       Intake of Radionuclides,  Health Physics, Vol. 19, pp. 187-222, 1970.

C-18   McClendon, J.F., Iodine and the Incidence ofGoiter, University of
       Minnesota Press, Minneapolis, 1939.

C-19   Bair, W.J., Plutonium Inhalation Studies,  Battelle Northwest Laboratory,
       BNWL 1221, 1970.

C-20   'International Commission on Radiological Protection, Publication No.6
       Pergamon Press, 1964,

C-21   Dolphin, G.W., The Biological Problems in the Radiological Protection
       of Workers Exposed to 23T>u,  Health Physics,  Vol. 20
       pp. 549-557, 1971.

C-22   Snyder, W., Internal Exposure,  Chapter 10, Principles of Radiation Pro-
       tection, K.Z. Morgan & J. Turner, John Wiley and Sons, 1967.

C-23   National Academy of Sciences - National Research Council, The Effects
       on Populations of Exposure to Low Levels of Ionizing Radiation,
       Report of the Advisory Committee on the Biological Effects of Ionizing
       Radiation (BEIR), U.S. Government Printing Office 1972.

C-24   Funk, F., Cytosine to Thymine Transitions from Decay of Cytosine-53H
       in Bacteriophage S 13,  Science, Vol. 166, pp. 1629-1631, 1969.

C-25   International Committee on Radiological Protection, Publication No.8
       Pergamon Press.

-------

-------
           APPENDIX D
Radiation and Economic Impact
     Analysis Techniques

-------

-------
                                  B-l




I.    INTRODUCTION




           This appendix presents the calculations! techniques uspd in




      arriving at radiation dose and health risk estimates for the general




      population due to nuclear fuel reprocessing and the general approach




      taken to assess the economic impact of effluent controls on the




      industry.  The dose and risk estimates are made for individuals




      living at 3 kilometers from a reprocessing plant,  the regional popu-




      lation within 80 kilometers (50 miles) of a reprocessing plant, for




      the total United States population (including that of the regional




      zone), and for the world population (excluding the U.S.  population),




      The dose and risk estimates are made for a representative single five




      tonne per day plant and for the projected industry as a  whole.  Only




      releases via the atmospheric route are considered  since  future plants




      are not expected to release significant quantities of radioactive




      material in liquid effluents.   One present facility has  significant




      liquid releases and this case should be considered on an individual




      basis.

-------
                              D-2
II.   ESTIMATED REGIONAL POPULATION EFFECTS


           The regional population around a reprocessing plant may be


      exposed to higher radioactive material concentrations due to releases


      from the plant than is the total U.S. population.   Thus, this


      regional population group is considered as a special case.  The total


      regional population dose received by a specific body organ (organ i)


      exposure to a specific radionuclide (radionuclide j) is estimated by


      using the following equation:




                D   =   0     (20          F.    C   ?„  person-rem/yr
                 i                         i         B.
                                        .
                              (3  km)     ij
   Where :
   Q.=  Release rate of radionuclide j from the fuel reprocessing plant


        in curies released per second (Ci/s) . .  The average annual


        release rate is determined by using the curie/tonne values


        in appendix A, the plant decontamination factors in appendix


        B, and the size of the plant in tonne/day . capacity .  For this


        evaluation, the capacity of a single plant is taken as


        5 tonne/day which is equivalent to 1500 tonne/year. . It


        is assumed that a plant operates at full capacity.


      = The meteorological dispersion factor, i.e., ttie radionuclide

(3 km)

        concentration in air, X, 3 km from the plant per unit of

                                            3
        release rate from the plant [(yCi/cm ) /(.Ci released/s) ] .


        A distance of 3 km is chosen as a reference point since the


        air concentration at this point is generally not significantly


        affected by the stack height of the plant.  Using the methods


        referenced in appendix C,  X/Q values have been calculated for

-------
                            D-3
      many sites by many people.  In this evaluation,  a value for


      (X/Q)3km) of 5 X 10~8 ( Ci/cm3)/(Ci/s) is considered to be


      representative of a typical plant.   This value is assumed to


      be uniform in all directions around the plant.  Actual plant


      values may vary from the above value by as much as factors of


      five or ten.


F   » The pathway dependent dose conversion factor which gives the


      dose to organ i due to a medium (air, food, or water) concen-

                                                   2
      tration of radionuclide j [(rem/year)/( Ci/cm )].  For


      example, iodine exposure of the thyroid gland by inhalation


      and milk ingestion would have two separate dose conversion


      factors.  The dose conversion factors used in this study are


      given in appendix C.  For radionuclides with long retention


      times in the body, e.g., the actinides, the conversion factor


      represents the equilibrium dose rate resulting from a continu-


      ous factor represents the equilibrium dose rate resulting


      from a continuous constant intake for several years.


C  «  The regional dilution and population distribution correction


      factor.  It is a ratio of the average individual dose within


      80 km of the plant to the average individual dose at 3 km.  It


      takes into account increased dilution as the radionuclides are


      transported further from the plant and an uneven distribution


      of population around the plant.  It can be calculated


      theoretically by assuming a population distribution (lef,


      D-l)  or it can be determined from population dose calculations


      around real plant sites,  for this study, the results of dose

-------
                          D-4
     calculations for about 50 reactor.sites (Ref«  D-2) were
     analyzed and a value of 0.028 (rem/person within 80  km)/
     rem/person at 3 km)  was obtained.   Theoretical results are
     similar.  This value was assumed to be the same for  all
     nuclides.   Fuel reprocessing plants are assumed to be located
     at sites with population characteristics similar to  those at
     reactor sites.  Individual plant correction factors  may vary
     by as much as a factor of five from the average value given
     above.  The distance of 80 km (50 miles) was chosen  as a cut-
     off on regional calculations since the distance is large enough

     to include any nearby large population center yet small enough
     so that the area affected can be considered a local  area.
R  = The populations within 80 km of the fuel reprocessing plant
     site.  The population values of the above-mentioned  50 reactor
     sites, taken primarily from environmental reports, lead to an
     average population around a site of 1.5 x 10  people in 1980.
     Population sizes around individual plants can vary from this
     by a factor of three.  The doubling time of this population
     is assumed to be about 40 years.  For purposes of characterizing
     age-specific factors, 2.5% of the population is considered to be
     under 1 year old, 45% between 1 and 20 years old, and the remainder
     over 20.  One-half of the population is used for genetically signi-
     ficant" dose calculations because of the child-bearing potential.
Using the'above factors,  the dose equation becomes:

-------
                             D-5
 D.. =   (Q  Ci/s) x

   J       J
     x (F.. ^7%) x (0.028) x (1.5 x 106 people)
         ij yCi/cm-3




     x (1.5 to take into account population growth)


     = 0.0032 Q. F   person-rem/year


     This represents the average annual population dose for a 40



year period of constant emissions during constant plant operation.
The equation can be modified further by considering that:


              A1 x 1500

              3.15 x 10? x d
              A1 x 1500         Ci/s
                            ,


where:



Aj = Curies of radionuclide j per tonne of fuel reprocessed



     (appendix A)



dj = Plant system decontamination factor for radionuclide j



     (appendix B) and 1500 tonne are processed per year at a 5



     tonne per day capacity, and there are 3.17 x 10  seconds per year)



Therefore:


                -7-L
  ^li ~ •'••-' x ^-0   
-------
                             D-6




where the J. values are those given in appendix C.   The total




regional health risk for one plant, H, is:
                    Hi =    (Ji    V
     The total annual fuel reprocessing industry effect on the



regional populations is obtained by multiplying the effect for



one plant by the number of operating plants in any given year



(appendix A).

-------
                                    D-7
III.  ESTIMATED EFFECTS OK THE UNITED ' STATES POPULATION



           Several radionuclid.es released in the gaseous effluents from



      a fuel reprocessing- plant spread from the local region to all or



      part of the total United States land area before being diluted



      around the world, . The method of estimating effects to the U.S.



      population depends on the radionuclide being, considered.  The radio-



      nuclides considered in this study are tritium, krypton-85, iodine-129,



      and the actinides.  Iodine-131 is considered to be only a regional



      problem because of its short half-life and the rate of local deposi-



      tion assumed.  Krypton-85 and tritium expose 'the U.S. population



      only briefly before subsequent dispersion throughout the world, while



      iodine-129 and the actinides are assumed to build up only on U.S. soil.



      In fact, the impact of these latter two may be restricted to a region.



      A.  Tritium



           It is assumed that most of the tritium released as a gaseous



      effluent from the reprocessing plant will enter into the hydrologic



      cycle.  The released tritium is assumed to fall out or rain out over



      the Eastern United States [3.7 x 10 6 km2 (1.5 x 10 6 mi )] and becomes



      diluted by the annual rainfall [40 inches (average 100 cm) ] (Ref>



      D-3) over this area.  With some further dilution by uncontaminated



      water, this then becomes the water concentration to which the popula-



      tion of the Eastern United States (80% of the total U.S. population)



      along with their animals and crops are exposed.



           The total population dose is then given by:
      D..  = (C.     )  x f  x (F..        )  x  0.80  P   person-rem/yr
       13      3 ml           ij  pCi/ml         u

-------
                              D-8
where:

C, - The water concentration of tritium determined by diluting the

     yearly input, Q. (Ci/s) x 3.15 x 10  s/yr, of tritium to

     the environment by the average annual rainfall over one-half

     of the United States.

I -  A factor to take into account dilution of tritium by uneontamin-

     ated water from deep artesian wells and the fact that not all

     tritium will fall out over the Eastern United States.  Some,

     perhaps a large portion, will fallout over the Atlantic Ocean

     and will be diluted in a larger volume.  For this evaluation

     f = 0.5 is used although it is only an estimate.

.. = Dose conversion factor from appendix G [100 (rem/yr)/(uCi/ml)

     for this case.]

   = Population of the United States.  The population growth of the

     U.S. is estimated as in figure D.I.  One-half P  is used for

     genetically significant dose calculations.

     Tritium concentration, C., is related to the environmental

input by;
     	3.15 x 107  s/yr	
     (1.5x10° miz)x(1.61x10^ cm/mi)%(40 in)x(2.54 cm/in)
     8,0 x 10~b Q pCi/ml


     2,5 x 10~13 -ji N pCi/ttl

-------
             D-9
                                                        Z
                                                        o
                                                        Cu

                                                        O
                                                        Q_

                                                        CO
                                                        oo

                                                        Q
                                                        LLJ

                                                        fc=

                                                        Z
                                                        D
                                                        Q
                                                        LU
                                                        I—

                                                        u
                                                        LU


                                                        O
                                                        Of.
                                                        Q_
                                                        3

                                                        O)
NOIXVindOd S3J.V1S QBIINH

-------
                             D-10



where:




 j = Curies/tonne - froto appendix A.



D. = System decontamination factor from appendix B.



N  = Number of tonnes processed per year.   (1500 tonne/yr for a



     single plant and as in appendix A for the total industry.)



Therefore:





                            'A.            person-rem/yr to U. S. A.

D,, = 2.5 x 10    x 0.80 f N —  F.,  P    population from the

 13                          d.   1J   U   fallout of tritium





"B.  Rrypton-85



     Part of the population of the Eastern United States is exposed



to air concentrations of krypton-85. as it passes from the fuel



reprocessing-plant  to'the  Atlantic  Ocean on its  first pass  around



the world.  The  dose from  this exposure  pathway  is  taken from a



study  recently performed at the  National Oceanic and Atmospheric



Administration  (Reference  D-4).   For  a plant in  Morris,  Illinois,



releasing one curie of'krypton-85 'per year,  the" population-weighted



concentration on its first pass  over  the Eastern United  States to


                               -16             3
the Atlantic Ocean  is"2.5  x 10    person-Ci/cm .   For purposes



of this  evaluation,•this value is considered adequate to use  for



all plants.  This value is then'multiplied  by total annual  releases



and dose conversion 'factors to obtain dose  values-.

-------
                             B-ll
                       person-_Ci
                                                         ±j     -






where N, A  , d . , and F. .  are  as  defined  previously.   The 1.5  factor
          «J   J         J

is to account for population  growth based  on  a  doubling  time  of  40


years .


Therefore :
                   ~10
                         A.
     D. . = 3.8 x 10 •"* N -L F. .
      ijj                 d.  ij
person-rem/yr to U. S,

population from release
                          -	1     of krypton-85
C. Iodine-129



     As a first approximation, all  of  the iodine-129  release is

                                                              /-    *\

assumed to fall out  over  the Eastern United States  [3.75 x 10 km


         fi    O
(1.5 x 10  mi) ] and uniformly mix with the stable iodine in  the



soil  to a depth of 20 cm.  This then becomes the specific activity



of iodine in the diet to which all persons in this part of the



country are exposed.i Because of its long half-life (1.6 x 10  years),



iodine-129 will build up on the soil and expose the population long



after it has been released to the environment.  The movement  of



iodine-129 in the biosphere is not well documented at the present



time.  However, even with the assumptions that must be made,  the



first estimate obtained is considered reasonable.

-------
                             B-12

      The  specific activity of  iodine-129 in the soil at time t,
    ,    129T/    127T   ,
 curies     I/gr     I,  is:
 Specific  _  Z H(f)  '  Aj  *; f
          -
Activity  -  o     d.          '    l>5  x 1Qb ^ x (]U(j x 1()5
            4  x 1°~6  S 127I  »  1.5  g soil    on    ,   ^
                     a - — •        •« — - —  x 20 cm depth
                     g soil      .3               r
                     °             cm


            -  2.2  x 10~13 I  N(f )  AJ  f  Ci  129I
                          o     d.       g 127i


where  £ is  the fraction  of  iodine-129  release that  stays in the

soil,  and it  also  includes  dilution  by iodine taken in from other

sources, and  the other factors are as  defined previously.   For

purposes of this evaluation,  f .is taken as 0.5.   This  value is  only

a rough estimate,  however,  and may be  considerably  in  error. Like-
                         —6  127
wise,  the value of 4  x 10   g     I/g soil  is  taken  from only one

Ref . (D-5) and may not be representative of the  Eastern United

States,  These numbers can  easily be changed  when better values

are determined.  Using the dose conversion factor given  in

appendix C, the annual thyroid population dose rate as a

function of time,  t,  is determined from:
                                r,129
     D   .- (specific activity  %r-^  x (4.4x10^         .... )
      ij                        g127!              Ci129I/g127I
            x  (0.80 P (t))
                       -8                   i
            =  7.7 x 10  P (t)  Z   I(f*)  —*•  f  thyroid person-rem/yr

-------
                             D-13



     This is the annual population dose  (thyroid) due to the total



amount of iodine-129 in the environment at the time of calculation.



For purposes of determining health risk, the iodine-129 is assumed



to expose the population for 100 years beyond the time of release.



D.  Actinides



   ,  The actinides are assumed to build up in the Eastern United



States in the same manner as iodine-129 with ..the principal ex-



posure pathway believed to be resuspension of the material and



inhalation.  The annual population dose to the lungs from the



buildup of actinides in the environment is;
            E    0.8 P (t) N(t')   l f R  F

            t'»o      U           d,    8
     D   .  	3

       J           3.8 x 10b  km2
                                                    3             2
where R  is the resuspension factor in.terms of Ci/m  air per Ci/m
       5


on the ground.  Based on calculations using fallout data and data


                                                           — H

from around Rocky Flats (Reference D-6), a value of R  - 1Q   is
                                                     S


used in this study.  The uncertainity.of this value may be of the



order of several orders of magnitude.   The fraction of actinides



released that remains on the soil is taken as 0.5 for this study.



Therefore:
D. ,
-19 t"
= 2.1 x 10 P (t)f I N(t')
0
A F
2> 3 .ij
J dj
person-rem
year

-------
                             D-14
The dose calculation is performed for plutonium-239 and the total


of all the long-lived actinides is about 10 times the plutonium-239


dose for long-term exposure.  This results from a consideration of A.
                                                                    J

and F.. values in appendixes A and C,  For purposes of determining


health risk, the actinides are assumed to expose the population for


100 years beyond the time of release.


     The health risks from all of these pathways are estimated by:
                  >..  J.
     Admittedly, the pathway models used in this section for the


determination of environmental effects to the United States popu-


lation are uncertain and unproven.  They are presented as a first


order approximation of these effects and the points of uncertainty,


especially relative to environmental behavior, are indicative of


research needs already identified in these areas.

-------
                                  D-15





IV.  EFFECTS ON THE WORLD POPULATION OUTSIDE THE U.S.




          Two radionuclides released in fuel reprocessing are expected to




     result in doses to the world population - krypton-85 and tritium.




     A.   Krypton-85




          The worldwide dose due to krypton-85 exposure can be estimated




     by diluting the output froift one year of fuel reprocessing into the




     world's atmosphere (5.14 x 10   g; sea  level air density » 0.00129


         o

     g/cm ) (Ref. D-7), and then determining population dose while it




     decays away.  The total yearly dose rate at any instant in time is




     a combination of contributions from all previous environmental




     releases of krypton-85.  The total population dose to be received




     by a specific body organ (organ i) over the total time of decay from




     a one year's input of krypton-85 is given by;





      t,  AJ    %   /  ^   ir,6  PCi s    .0.00129 g/cm3         rem/yr
      Dljr (decay) - <— x 10   ^= ) x  (5<14 x ^L g) * (F.   C1/m3) x
        3





            / Pw(t) x e~xt'dt'"




     where:




   X    - Decay constant for krypton-85 (0.0645/yr)




 P (t)' - World population at time t; (t-0 at time of release).  The



                                                9
          world population is taken as 3.56 x 10  people in 1970 with




          an annual growth rate of 1.9%/yr (Ref. D-8).  For genetically




          significant dose calculations, one-half the world population




          is used.  Five percent of the world effect is subtracted to




          account for the United States contribution to world population dose,




          The other factors are as defined previously.

-------
                             D-16
 Therefore:
D£. (decay) = 1.9 x l(
                       8
                                      .  e
                                        °*019t
tdtal wbrld persott~fem
contaitted by one year's
release (less U^ S, dose),
where t" is measured from 1970 to  the  time of release.

     The annual dose rate at time  t from  the buildup of krypton-85

in the atmosphere from releases i's given  bys
D  (buildup) =
                             .    1A6    -0.0645(t-f)
                            j  x 10  x e        s     '
                    o   d.
               x
                       [
                       O.OQ129
                       5.14 x 1021 g
                                   ij    w
where t=t'=0 in 1970.
Thus:  D.. (buildup) - 8.5 x lO'10   -  F. ,-°-0455tx Z N(tOe*°*°645t'
person-rem/yt to the world's population  (less U
-------
                             B-17




                                              19
 the  Northern Hemisphere (one-half of 2.7 x 10   1) (Ref, D-7)



 and  assuming that the Northern Hemisphere's population (80% of the



 world  population) (Ref. D-8) is exposed to the resulting



 concentration.   The U.S.  contribution (about 7%) is subtracted out.



     The  total population dose for organ i that will be received



 over the  total time of decay from a one year's input of tritium



 is given by:






                    NA.     6   ci              2
     D  . (decay) - (—^ x  10  ^~)  x  (	—	)

     • 1J              j         y      2.7 x  10xy  x  10J ml




                                    r „  ,.,,   -0.0562t*   *
                             NA
     D, .(decay) '- 5.3 x 10"6 -J- F.. e0'019t
                              Q
                                .
world person-rem

committed
by one year's release of tritium  (less U. S. dose) where  t  is



measured from 1970 to the time of release.



     The annual dose -rate at time t from the buildup of U.  S.  nuclear



power generated tritium released to the environment is given by:
     n  A. -1J  N           j      H  -Q.0562(t-t")
     D.. (buildup) =* Z —»-:	J-  x 10  e

      13
                    x  f	2	> x F   x P  (t)
                       V*        ntj   J    "f -|    T£r

                       2.7 x 1022 ml      3
where:  t = t» = 0 in 1970,

-------
                             D-18




     DI. (buildup) - 0.20 x Hf6  f- I   e-°'0372t Z N(t') e40'0562'"

                                  j   J           o  '



person-rem/yr to world's population from U. §. reactor produced


tritium (less U. S. dose) .  For genetically significant doses use


one-half this value.



     Again,, to evaluate future doses due to the environmental

                                  -0.0562T   0,0191
buildup of tritium, multiply .by e       '.  e       and integrate


from T==o to T-«> to account for population growth and decay.


    ; The health effects, resulting, from  exposure to tritium and



krypton-85 are determined by:



               H = V" D • . J*
                   2-  ^  i
(Some radionuclide fuel content and system decontamination


factors used in these calculations are given in table D,l)

-------
                    D-19
                   Table D.I

Factors Used in the Assessment of Environmental
   Radiological Impact of Fuel Reprocessing
H-3
Kr-85
1-129
1-131
Actinides
Activity
in Fuel
(Ci/MT)
800.00
10,500.00
0.04
2.00
(See Appendix A
Plant
Decontamination
Factor
1.0
1.0
1000
1000
•••• io9 •
          fable A.4)

-------
                                  D-20





V.   ECONOMIC ANALYSES




          Annual-costs should include considerations of debt service,




     depreciation, and Federal taxes.  The first twd are dependent on




     the assumed plant life, salvage value, investment capital and




     debt.  An example of the total annual costs for a 5 tonne/day




     plant is presented in table D-2 (Ref. D-9).  In this regard, the




     total economic impact upon the investors of the inclusion of




     a particular control system requires consideration of the cost of




     money invested over the life of the plant.  This, of course, will



     be somewhat offset by the increased value of the product over




     the same period.




          Several methods are available for presenting the cost ascrib-




     able solely to control systems.  These methods are described as




     annual cost, present worth, and total commitment.




          In order to obtain estimates of these values, it is necessary




     to specify first costs (P), interest rate  (i), debt lifetime (n)




     plant life and estimated salvage value (L).  In addition a discount




     factor of 7.5% was used to estimate present worth.  For the present




     analysis, the lifetime is considered to be 40 years, the salvage value




     is considered negligible, and the effective interest rate is assumed to




     be either 10% or 24%.  These estimates of costs are based upon the




     assumption that the control systems are add-on facilities and not directly




     required for processing.




          The annual costs are considered to be the sum of capital




     recovery costs and the annual operating expenses  (A.O.E.).  Annual




     costs therefore are estimated from the following  equation:




          Annual cost = P x erf -f A.O.I,

-------
                                                                   TABLE D.2




                                         Estimate of the Economics of a 5 MTU/Day Reprocessing Plant—7
(Note that "000" is omitted from all


Tear '
1971 -2
1972 -1
1973 0
1974 1
1975 2
1976 3
1977 4
1978 5
1979 6
1980 7
1981 8
1982 9
1983 10' •
1984 11
1985 12
1986 13
1987 14
1988 15
Fuel
Processed,
Metric
Tons /Year



290
690
1,200
1,320
1,500
1,500
1,500
, 1,500
1,500
1,500
- '1,500
1..500
: 1,500
1,500
1,500


Operating
Costs



7,708
9,975
12,070
12,520
13,215
13,215
13,215
13,215
13,215
13,215
13,215
13,215
13,215
13,215-
13,215


Debt
Service



2,567
2,567
2,567
2,567
2,567
2,567
2,567
2,567
2,567
2,567 -
2,567
2,567
2,567
2,567
2,567


Depre- ,
elation—



9,750
9,130
8,420
7,800
7,170
6,500
5,840
5,220
4,550 '
3,500
3,270
2,570
1,950"
1,330
620

Other
Fixed ,
Co-sts5-'



2,430
2,430
2,430
2,430
2,430
2,430
2,430
2,430
2,430
2,430 '
2,430
2,430
2,430
2,430
2,430


Total
Costs



22,455
24,102
25,487
25,317
25,382
24,712
24,052
23,432
22,762
22,112
21,482
20,782
20,162
19,542
18,832
money values)
Gross
Receipts
at
$27.50/kg



7,975
19,000
33,000
36,300
41,250
41,250
41,250
41,250
41,250
41,250
41,250
41,250
41,250
41,250
41,250


Before-
Tax Income



(14,480)
(5,120)
7,513
10,983
15,868
16,538
17,198
17,818
18,468
19,138
19,768
20,468
21,088
21,708
22,418


Federal
fax4'







7,090
7,940
8,250
8,560
8,860
9,200
9,480
9,820
10,100
10,400
10,780


Cash Flow
(18,200)
(18,200)
(18,200)
(4,730)
4,028
15,933
18,783
15,948
15,098
14,788
14,478
14,158
13,838
13,558
13,218
12,938
12,638
12,258


Obliga-
tion
-19,660
-42,470
-68,870
-85,370
-94,342
-90,939
-83,685
-78,568
-73,590
-68,202 o
-62,324 N>
-55,856 *"*
. -48,748
-40,820
-32,022
-22,134
-11,016
+1,242
— 70 percent equity at 16 percent;  30 percent  debt  at  7  percent.



— Sum-of-year digits.


c/
— State ^nd local taxes  (2.51 percent),-'insurance  (0.25  percent),  and  interim replacement  (0.3,5 -percent).



— At 48 percent.




       From - Lecture VIJ. - Nuclear Fuel  Processing, G.G.  Eichhole,  1972 Short Course at Georgia  Tech.,  re: Nuclear Fuels-Management and Economics

-------
                                   D-22






where P and A.O.E. are obtained from the estimates presented in table




B-2.  The capital recovery factor  (erf) for 40 years at 10% is equal




to  .105 (Ref. D-10).




    Annual costs can be converted  to present worth of all




disbursements for lifetime, n.  This is accomplished by multiplication




of the annual costs by the present worth factor  (pwf) for the




appropriate lifetime and interest rate and by adding the result to the




first cost;  PW = A.C. (pwf) + P.




    The present worth factor for 40 years at 10% is 9.779 (Ref. D-10).




    The total control system cost  can also be estimated over the life




of the plant.  Assuming that the principal cost is repaid within 10




years at an interest rate of 10%,  the total cost over 40 years is




estimated to be as follows:




      T.C. = (P x erf 1 ) x 10 + A.O.E. x 40




where:  erf  = .163



The system cost presented in table D-3 is in terms of total cost for




both 10% interest with 10 yr debt life and 24% interest with 20 yr debt




life.  In addition, the present worth for the latter parameters is




presented.  These figures do not consider the loss of opportunity




costs.




    The costs to the investors in  time and money associated with




litigation because of suits by opponents are not presently known, but




could be substantial.  Introduction of expensive control systems




during the initial construction stage and early use might be cost-




effective in an overall view even  if the system might not be cost-




effective on the basis of risk reduction alone.

-------
                                           TABLE D.3

                                     CONTROL SYSTEM COSTS
                                    First Cost
                                      10% Int.
                                      10 yr. debt
                                      40 yr. life
                                      Total Cost
                           24% eff. int.
                           20 yr. debt
                           40 yr. life
                           Total Cost
                          241 eff. int.
                          20 yr. debt
                          7.5% discount
                          40 yr. life
                             ..P. W.
Isotope
Krypton-85


Iodines
Actinides

System (10 Dollar)
•. Cryogenic
•'• Distillation - 3
' Cryogenic Adsorption 3
Freon-Adsorption 1.5
; Caustic Scrubber
and AgZ Adsorption 1.2
HEPA Filter (2) .1
HEPA + Sand .35
(10 Dollar)
8.9
11.1
6.4
6
2.2
'• 3.3
(10 Dollar)
19
21
11
10
2.5
4.5
(10 Dollar)
11.5
12.1
6.4
5.4
1
2.1



*


                                                                                                       e
                                                                                                       to
Tritium
Voloxidation
10
20
54
31
   The present worth of a caustic scrubber alone is estimated to he  $3*85 x 10 ,

-------
                                   D-24



    Beyond the normal costs considered, there is also the matter that



failure to introduce control systems for effluents that could have



worldwide distribution (tritium and krypton-85) could be costly to the



Federal government in terms of foreign policy expenditures,  The



magnitude of these costs is unclear at present, however.



    The impact of fuel reprocessing plants upon the  economy, e.g., on



industries which depend upon the economy, e.g., on industries which



depend upon the use of radiosensitive materials has been estimated to



be negligible as long as the media concentration of radioisotopes are



lower than the constraining levels set for humans by ICRP (Ref. B-ll,



D-12),



    Cost Effectiveness Considerations.  The effectiveness of risk



reduction of the previously described isotope specific control



(appendiK B) is dependent upon: (1) the group at risk which is



considered, and (2) the specific health effects used as a risk index.



The rate of decrease in number of health effects avoided by sequential



addition of the individual effluent control systems is dependent upon



the orcfef in which the systems are incorporated.  The number of health



effects avoided by incorporation of the individual systems in the



order of decreasing unit cost-effectiveness is presented as a function



of the cumulative system control cost in figures D-2 and D-3.  The



individual component for control of particulates, iodines, krypton,


                                                           3    3 •
and tritium were assumed to have decontamination factors 10 » 10 ,


  2        2
10 , and 10   with present worth costs of 1, 5.4, 6.4, and 31 million



dollars respectively.  A system capable of removing  the residual



pollutants was assumed to be available with a cost of 20 million

-------
                                   D-25



dollars also.  Figure D-2 displays the reduction of the projected




risks of mortality (cancer induction and genetically coupled) as a




function of cumulative system costs for the regional, Eastern United




States, and worldwide population groups.  The health risks considered




and presented in figure D-3 include nonfatal cancer induction as well.




The figures show that in every instance, the particulate control




system is the most cost-effective.  The iodine control system is the




next most cost~effective, except for the situation which considers the




estimated mortalities to the world population which are avoided.  When




considering the world population group, the iodine and krypton control




systems appear to be practically equally cost-effective for risk




reduction.  In consideration of the regional and United States




population groups, tritium control appears the next most cost-




effective to the particulate and iodine systems although krypton




control is not significantly different from that of the tritium




control system.




    The cost-effectiveness of a combined system for any group at risk




can be estimated from the slope of a directed line segment which




connects the origin and the appropriate arrow tip representing the




combination.  The steeper the slope, the greater the cost-




effectiveness.  It can be seen that the cost-effective excess of a




combined system will be necessarily less than that of its most cost-




effective component,

-------
                                   D-26
a
LU
Q
O
t/}
t—
u
UJ
u_
u.
LU
   1400-
   1200-
   1000-
800-
600-
    400-
    200
      0
               FIGURE D - 2  RJSK REDUCTiON(MORTALITY AND MORBIDITY)
                          vs CUMULATIVE CONTROL SYSTEM COST
              NUCLEAR FUEL REPROCESSING - 40 YEAR OPERATION
                      5 TONNES/DAY  THROUGHPUT

        COMPONENT CONTROL SYSTEMS
        P - PARTICULATE   T - TRITIUM
        I-IODINE       R-RESIDUAL
        K - KRYPTON
                                 Total World Population
U.S. Population (excluding regional)
                 _         R
         K
                            Regional Population (only)
                                    (50 miles)
                            COMPONENT CONTROL SYSTEMS
                               ADDED IN THE ORDER OF
                            DECREASING COST EFFECTIVENESS
VP
o
1
10
1
20
i
30
i
40
I
50
I
60 7(
                     CUMULATIVE  CONTROL SYSTEM COST
                     :               (106 Dollars)

-------
                                B-27
        FIGURE D - 3  RISK REDUCTION (MORTALITY) vs CUMULATIVE CONTROL SYSTEM COST
LU
P
Og
UJ
 800-
600-
 400-
 200-
   0.
              NUCLEAR FUEL REPROCESSING - 40 YEAR OPERATION
                      5 TONNES/DAY THROUGHPUT

        COMPONENT CONTROL SYSTEMS
         P - PARTICULATE   T - TRITIUM
         I-IODINE       R-RESIDUAL
         K - KRYPTON
            COMPONENT CONTROL SYSTEMS ADDED IN THE      OF
              DECREASING COST EFFECTIVENESS
                              Total World Population
         K
                       U.S. Population (excluding regional)
         i*"
                                                   R-
                            Regional Populational(only)
                                     (50 miles)
     0
              10       20       30       40       50       60
                  CUMULATIVE CONTROL SYSTEM COST
                                (106 Dollars)
70

-------
                                   D-28
                              REFERENCES
D-l  Knox, J.B.  Airborne Radiation from the Nuclear Power Industry,
     Nuclear News, 14:  27-32, February 1971.

D-2  Gamertsfelder, G.C.  Statement on the Selection of As Low As
     Practical Design Objectives and Technical Speciations for
     the Operation of Light Water Cooled Nuclear Power Reactors,
     Presented at AEC Hearings on the "As Low As Practical Concept, "
     1972.

D-3  Statistical Abstract of the United States, 1969,.U.S. Department
     of Commerce.

D-4  Machta, L. National Oceanic and Atmospheric Administration,
     Unpublished data.

D-5  Wayne, I,J., Demetrios A. Kontros, and W.D. Alexander.   Clinical
     Aspects of Iodine Metabolism, F.A. Davis Co., Philadelphia, Pa.
     1964.

D-6  Compendium of Environmental Surveillance Around the Rocky Flats
     Plutonium Plant, report prepared by FOD/ORP/EPA, November 1972.

D-7  Klement, A.W., Jr., C.R, Miller, R.P. Minx, and B. Sttleien.
     Estimates of Ionizing Radiation Doses in the United States -
     1960-2000, U.S. EPA document, CSD/ORP 72-1, 1972.

D-8  United Nations Statistical Office Report, 1966.

D-9  G.G. Eicholz.  Lecture VII - Nuclear Fuel Reprocessing  - 1972,
     Short course re nuclear fuels, management and economics,
     Georgia Tech.

D-10 Grant, E.I. and W. Ireson.  Principles of Engineering Economy.

D-ll Frieser, H., G. Heiman, and E, Rane.  The Effect of Radioactive
     Nuclides of Photographic Emulsions, Photographische Korrespondenz,
     98,  131-140, 1962.

B-12 Clear, Murray.  Personal communication, Eastman Kodak Co.,
     Rochester, New York.
                                            AU,S, GOVERNMENT PRINTING OFFICE: 1973 546-311/111 1-3

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U.S. ENVIRONMENTAL PROTECTION AGENCY
        Washington, O.C. 20460
         OFFICIAL BUSINESS
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U.S. ENVIRONMENTAL PROTECTION
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           EPA-335
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